METHOD OF APPLYING NICKEL COATINGS ON URANIUM

DOEpatents

Gray, A.G.

1959-07-14

A method is presented for protectively coating uranium which comprises etching the uranium in an aqueous etching solution containing chloride ions, electroplating a coating of nickel on the etched uranium and heating the nickel plated uranium by immersion thereof in a molten bath composed of a material selected from the group consisting of sodium chloride, potassium chloride, lithium chloride, and mixtures thereof, maintained at a temperature of between 700 and 800 deg C, for a time sufficient to alloy the nickel and uranium and form an integral protective coating of corrosion-resistant uranium-nickel alloy.

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

DOEpatents

Travelli, A.

1985-10-25

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

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

DOEpatents

Travelli, Armando

1988-01-01

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

Method of making crack-free zirconium hydride

DOEpatents

Sullivan, Richard W.

1980-01-01

Crack-free hydrides of zirconium and zirconium-uranium alloys are produced by alloying the zirconium or zirconium-uranium alloy with beryllium, or nickel, or beryllium and scandium, or nickel and scandium, or beryllium and nickel, or beryllium, nickel and scandium and thereafter hydriding.

Interaction Between U-Mo Alloys and Alloys Al-Be

NASA Astrophysics Data System (ADS)

Nikitin, S. N.; Tarasov, B. A.; Shornikov, D. P.

The main objective of the work is the experimental determination of the effect of doping on the kinetics of the interaction of beryllium, aluminum and uranium-molybdenum alloy dispersed in the nuclear fuel. It is shown that an increase in the content of Be in Al leads to a linear decrease in the rate of interaction of the alloy with uranium-molybdenum alloy. Besides AlBe-alloys have higher thermal and mechanical properties than other matrix alloys such as AlSi.

JACKETING URANIUM

DOEpatents

Saller, H.A.; Keeler, J.R.

1959-07-14

The bonding to uranium of sheathing of iron or cobalt, or nickel, or alloys thereof is described. The bonding is accomplished by electro-depositing both surfaces to be joined with a coating of silver and amalgamating or alloying the silver layer with mercury or indium. Then the silver alloy is homogenized by exerting pressure on an assembly of the uranium core and the metal jacket, reducing the area of assembly and heating the assembly to homogenize by diffusion.

URANIUM COMPOSITIONS

DOEpatents

Allen, N.P.; Grogan, J.D.

1959-05-12

This patent relates to high purity uranium alloys characterized by improved stability to thermal cycling and low thermal neutron absorption. The high purity uranium alloy contains less than 0.1 per cent by weight in total amount of any ore or more of the elements such as aluminum, silicon, phosphorous, tin, lead, bismuth, niobium, and zinc.

METAL COMPOSITIONS

DOEpatents

Seybolt, A.U.

1959-02-01

Alloys of uranium which are strong, hard, and machinable are presented, These alloys of uranium contain bctween 0.1 to 5.0% by weight of at least one noble metal such as rhodium, palladium, and gold. The alloys may be heat treated to obtain a product with iniproved tensile and compression strengths,

URANIUM ALLOYS

DOEpatents

Seybolt, A.U.

1958-04-15

Uranium alloys containing from 0.1 to 10% by weight, but preferably at least 5%, of either zirconium, niobium, or molybdenum exhibit highly desirable nuclear and structural properties which may be improved by heating the alloy to about 900 d C for an extended period of time and then rapidly quenching it.

Process for recovering niobium from uranium-niobium alloys

DOEpatents

Wallace, Steven A.; Creech, Edward T.; Northcutt, Walter G.

1983-01-01

Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and leave an insoluble residue of niobium stannide, then separating the niobium stannide from the acid.

Process for recovering niobium from uranium-niobium alloys

DOEpatents

Wallace, S.A.; Creech, E.T.; Northcutt, W.G.

1982-09-27

Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and form a precipitate of niobium stannide, then separating the precipitate from the acid.

Method for fabricating uranium alloy articles without shape memory effects

DOEpatents

Banker, John G.

1985-01-01

Uranium-rich niobium and niobium-zirconium alloys possess a characteristic known as shape memory effect wherein shaped articles of these alloys recover their original shape when heated. The present invention circumvents this memory behavior by forming the alloys into the desired configuration at elevated temperatures with "cold" matched dies and maintaining the shaped articles between the dies until the articles cool to ambient temperature.

Method for fabricating uranium alloy articles without shape memory effects

DOEpatents

Banker, J.G.

1980-05-21

Uranium-rich niobium and niobium-zirconium alloys possess a characteristic known as shape memory effect wherein shaped articles of these alloys recover their original shape when heated. The present invention circumvents this memory behavior by forming the alloys into the desired configuration at elevated temperatures with cold matched dies and maintaining the shaped articles between the dies until the articles cool to ambient temperature.

TERNARY ALLOY-CONTAINING PLUTONIUM

DOEpatents

Waber, J.T.

1960-02-23

Ternary alloys of uranium and plutonium containing as the third element either molybdenum or zirconium are reported. Such alloys are particularly useful as reactor fuels in fast breeder reactors. The alloy contains from 2 to 25 at.% of molybdenum or zirconium, the balance being a combination of uranium and plutonium in the ratio of from 1 to 9 atoms of uranlum for each atom of plutonium. These alloys are prepared by melting the constituent elements, treating them at an elevated temperature for homogenization, and cooling them to room temperature, the rate of cooling varying with the oomposition and the desired phase structure. The preferred embodiment contains 12 to 25 at.% of molybdenum and is treated by quenching to obtain a body centered cubic crystal structure. The most important advantage of these alloys over prior binary alloys of both plutonium and uranium is the lack of cracking during casting and their ready machinability.

Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

2014-07-01

The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

PROCESS OF DISSOLVING ZIRCONIUM ALLOYS

DOEpatents

Shor, R.S.; Vogler, S.

1958-01-21

A process is described for dissolving binary zirconium-uranium alloys where the uranium content is about 2%. In prior dissolution procedures for these alloys, an oxidizing agent was added to prevent the precipitation of uranium tetrafluoride. In the present method complete dissolution is accomplished without the use of the oxidizing agent by using only the stoichiometric amount or slight excess of HF required by the zirconium. The concentration of the acid may range from 2M to 10M and the dissolution is advatageously carried out at a temperature of 80 deg C.

REGENERATION OF FISSION-PRODUCT-CONTAINING MAGNESIUM-THORIUM ALLOYS

DOEpatents

Chiotti, P.

1964-02-01

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

FABRICATION OF URANIUM-ALUMINUM ALLOYS

DOEpatents

Saller, H.A.

1959-12-15

A process is presented for producing a workable article of a uranium- aluminum alloy in which the uranium content is between 14 and 70% by weight; aluminum powder and powdered UAl/sub 2/, UAl/sub 3/, UAl/sub 5/, or UBe/sub 9/ are mixed, and the mixture is compressed into the shape desired and sintered at between 450 and 600 deg C.

Distillation of cadmium from uranium plutonium cadmium alloy

NASA Astrophysics Data System (ADS)

Kato, Tetsuya; Iizuka, Masatoshi; Inoue, Tadashi; Iwai, Takashi; Arai, Yasuo

2005-04-01

Uranium-plutonium alloy was prepared by distillation of cadmium from U-Pu-Cd ternary alloy. The initial ternary alloy contained 2.9 wt% U and 8.7 wt% Pu other than Cd, which were recovered by molten salt electrolysis with liquid Cd cathode. The distillation experiments were conducted in 10 g scale of the initial alloy using a small-scale distillation furnace equipped with an evaporator and a condenser in a vacuum vessel. After distillation at 1073 K, the weight of the residue was in good agreement with that of the loaded actinides, where the content of Cd decreased to less than 0.05 wt%. The uranium-plutonium alloy product was recovered without adhering to the yttria crucible. The cross section of the product was observed using electron probe micro-analyzer and it was found to consist of a dense material. Almost all of the evaporated Cd was recovered in the condenser and so enclosed well in the apparatus.

METHOD OF DISSOLVING REFRACTORY ALLOYS

DOEpatents

Helton, D.M.; Savolainen, J.K.

1963-04-23

This patent relates to the dissolution of alloys of uranium with zirconium, thorium, molybdenum, or niobium. The alloy is contacted with an anhydrous solution of mercuric chloride in a low-molecular-weight monohydric alcohol to produce a mercury-containing alcohol slurry. The slurry is then converted to an aqueous system by adding water and driving off the alcohol. The resulting aqueous slurry is electrolyzed in the presence of a mercury cathode to remove the mercury and produce a uranium-bearing aqueous solution. This process is useful for dissolving irradiated nuclear reactor fuels for radiochemical reprocessing by solvent extraction. In addition, zirconium-alloy cladding is selectively removed from uranium dioxide fuel compacts by this means. (AEC)

Carbon-free induction furnace

DOEpatents

Holcombe, Cressie E.; Masters, David R.; Pfeiler, William A.

1985-01-01

An induction furnace for melting and casting highly pure metals and alloys such as uranium and uranium alloys in such a manner as to minimize contamination of the melt by carbon derived from the materials and the environment within the furnace. The subject furnace is constructed of carbon free materials and is housed within a conventional vacuum chamber. The furnace comprises a ceramic oxide crucible for holding the charge of metal or alloy. The heating of the crucible is achieved by a plasma-sprayed tungsten susceptor surrounding the crucible which, in turn, is heated by an RF induction coil separated from the susceptor by a cylinder of inorganic insulation. The furnace of the present invention is capable of being rapidly cycled from ambient temperatures to about 1650.degree. C. for effectively melting uranium and uranium alloys without the attendant carbon contamination problems previously encountered when using carbon-bearing furnace materials.

Non-carbon induction furnace

DOEpatents

Holcombe, C.E.; Masters, D.R.; Pfeiler, W.A.

1984-01-06

The present invention is directed to an induction furnace for melting and casting highly pure metals and alloys such as uranium and uranium alloys in such a manner as to minimize contamination of the melt by carbon derived from the materials and the environment within the furnace. The subject furnace is constructed of non-carbon materials and is housed within a conventional vacuum chamber. The furnace comprises a ceramic oxide crucible for holding the charge of metal or alloys. The heating of the crucible is achieved by a plasma-sprayed tungsten susceptor surrounding the crucible which, in turn, is heated by an rf induction coil separated from the susceptor by a cylinder of inorganic insulation. The furnace of the present invention is capable of being rapidly cycled from ambient temperatures to about 1650/sup 0/C for effectively melting uranium and uranium alloys without the attendant carbon contamination problems previously encountered when using carbon-bearing furnace materials.

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.

ALLOY COATINGS AND METHOD OF APPLYING

DOEpatents

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

1958-08-26

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

Process for massively hydriding zirconium--uranium fuel elements

DOEpatents

Katz, N.H.

1973-12-01

A method is described of hydriding uranium-zirconium alloy by heating the alloy in a vacuum, introducing hydrogen and maintaining an elevated temperature until occurrence of the beta--delta phase transformation and isobarically cooling the composition. (Official Gazette)

PLURAL METALLIC COATINGS ON URANIUM AND METHOD OF APPLYING SAME

DOEpatents

Gray, A.G.

1958-09-16

A method is described of applying protective coatings to uranlum articles. It consists in applying chromium plating to such uranium articles by electrolysis in a chromic acid bath and subsequently applying, to this minum containing alloy. This aluminum contalning alloy (for example one of aluminum and silicon) may then be used as a bonding alloy between the chromized surface and an aluminum can.

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.

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, John P.; Miller, William E.

1989-01-01

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

ELECTROLYSIS OF THORIUM AND URANIUM

DOEpatents

Hansen, W.N.

1960-09-01

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

Some observations on uranium carbide alloy/tungsten compatibility

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1972-01-01

Chemical compatibility between both pure and thoriated tungsten and uranium carbide alloys was studied at 1800 C for up to 3300 hours. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, dependent upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. The presence of a thermal gradient had no effect on the reactions observed nor did the presence of thoria in the tungsten clad.

Some observations on uranium carbide alloy/tungsten compatibility.

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1972-01-01

Results of chemical compatibility tests between both pure tungsten and thoriated tungsten run at 1800 C for up to 3300 hours with uranium carbide alloys. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, depending upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. Neither the presence of a thermal gradient nor the presence of thoria in the tungsten clad affect the reactions observed.

PURIFICATION OF URANIUM FUELS

DOEpatents

Niedrach, L.W.; Glamm, A.C.

1959-09-01

An electrolytic process of refining or decontaminating uranium is presented. The impure uranium is made the anode of an electrolytic cell. The molten salt electrolyte of this cell comprises a uranium halide such as UF/sub 4/ or UCl/sub 3/ and an alkaline earth metal halide such as CaCl/sub 2/, BaF/sub 2/, or BaCl/sub 2/. The cathode of the cell is a metal such as Mn, Cr, Co, Fe, or Ni which forms a low melting eutectic with U. The cell is operated at a temperature below the melting point of U. In operation the electrodeposited uranium becomes alloyed with the metal of the cathode, and the low melting alloy thus formed drips from the cathode.

METHOD AND FLUX COMPOSITION FOR TREATING URANIUM

DOEpatents

Foote, F.

1958-08-23

ABS>A flux composition is described fer use with molten uranium or uranium alloys. The flux consists of about 46 weight per cent calcium fiuoride, 46 weight per cent magnesium fluoride and about 8 weight per cent of uranium tetrafiuoride.

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

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

McDeavitt, Sean M

2011-04-29

Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºCmore » to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow, entitled “Uranium Metal Powder Production, Particle Distribution Analysis, and Reaction Rate Studies of a Hydride-Dehydride Process"« less

NICKEL COATED URANIUM ARTICLE

DOEpatents

Gray, A.G.

1958-10-01

Nickel coatings on uranium and various methods of obtaining such coatings are described. Specifically disclosed are such nickel or nickel alloy layers as barriers between uranium and aluminum- silicon, chromium, or copper coatings.

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.

PROCESS FOR DISSOLVING BINARY URANIUM-ZIRCONIUM OR ZIRCONIUM-BASE ALLOYS

DOEpatents

Jonke, A.A.; Barghusen, J.J.; Levitz, N.M.

1962-08-14

A process of dissolving uranium-- zirconium and zircaloy alloys, e.g. jackets of fuel elements, with an anhydrous hydrogen fluoride containing from 10 to 32% by weight of hydrogen chloride at between 400 and 450 deg C., preferably while in contact with a fluidized inert powder, such as calcium fluoride is described. (AEC)

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)

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)

Theoretical Model for Volume Fraction of UC, 235U Enrichment, and Effective Density of Final U 10Mo Alloy

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

Devaraj, Arun; Prabhakaran, Ramprashad; Joshi, Vineet V.

2016-04-12

The purpose of this document is to provide a theoretical framework for (1) estimating uranium carbide (UC) volume fraction in a final alloy of uranium with 10 weight percent molybdenum (U-10Mo) as a function of final alloy carbon concentration, and (2) estimating effective 235U enrichment in the U-10Mo matrix after accounting for loss of 235U in forming UC. This report will also serve as a theoretical baseline for effective density of as-cast low-enriched U-10Mo alloy. Therefore, this report will serve as the baseline for quality control of final alloy carbon content

Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

DOEpatents

McLean, II, William; Miller, Philip E.

1997-01-01

A method for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction.

Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

DOEpatents

McLean, W. II; Miller, P.E.

1997-12-16

A method is described for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction. 3 figs.

Characterization of microstructural, mechanical and thermophysical properties of Th-52U alloy

NASA Astrophysics Data System (ADS)

Das, Santanu; Kaity, S.; Kumar, R.; Banerjee, J.; Roy, S. B.; Chaudhari, G. P.; Daniel, B. S. S.

2016-11-01

Th-52 wt.% U alloy has a microstructure featuring interspersed networks of uranium rich and thorium rich phases. Room temperature hardness of the alloy is more than twice that of unalloyed thorium. The alloy age hardens (550 °C) only slightly (peak hardness/hardness of solution heated and quenched = 1.05). Room temperature thermal conductivity (25.6 W m-1 °C-1) is close to that of uranium and most of the binary and ternary metallic alloy fuel materials. Average linear coefficient of thermal expansion (CTE) of Th-52 wt.% U alloy [11.2 × 10-06 °C-1 (27-290 °C) and 16.75 × 10-06 °C-1 (27-600 °C)] are comparable with that of many metallic alloy fuel candidates. Th-52 wt.% U alloy with non-age hardenable microstructure, appreciable thermal conductivity, moderate thermal expansion may find metallic fuel applications in nuclear reactors.

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.

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

Rabin, S.A.; Lotts, A.L.; Hammond, J.P.

Uranium --molybdenum alloy rods containing from 10 to 15 wt% Mo and 1/16- in. in diameter were successfully fabricated by hot rotary swaging, followed by machining to remove the protective sheathing (Inconel with molybdenum barrier). Structurally strong rods with densities greater than 95% of theoretical were produced from both calciumreduced uranium mixed with hydrogen-reduced molybdenum and acid-cleaned, prealloyed shot when reduced in area about 55% at 1050 or 1100 deg C. Alloy homogeneity was good with prealloyed powders; however, traces of molybdenum -rich, gamma phase persisted in the elemental uranium -molybdenum material after swaging at 1100 deg C. Swagings embodyingmore » hydride uranium or oxide- contaminated prealloyed shot were unsatisfactory because of insufficient consolidation or poor interparticle bonding. (auth)« less

Nuclear fuel alloys or mixtures and method of making thereof

DOEpatents

Mariani, Robert Dominick; Porter, Douglas Lloyd

2016-04-05

Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

Uranium chloride extraction of transuranium elements from LWR fuel

DOEpatents

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

1992-08-25

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

Uranium chloride extraction of transuranium elements from LWR fuel

DOEpatents

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

1992-01-01

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

Intergranular tellurium cracking of nickel-based alloys in molten Li, Be, Th, U/F salt mixture

NASA Astrophysics Data System (ADS)

Ignatiev, Victor; Surenkov, Alexander; Gnidoy, Ivan; Kulakov, Alexander; Uglov, Vadim; Vasiliev, Alexander; Presniakov, Mikhail

2013-09-01

In Russia, R&D on Molten Salt Reactor (MSR) are concentrated now on fast/intermediate spectrum concepts which were recognized as long term alternative to solid fueled fast reactors due to their attractive features: strong negative feedback coefficients, easy in-service inspection, and simplified fuel cycle. For high-temperature MSR corrosion of the metallic container alloy in primary circuit is the primary concern. Key problem receiving current attention include surface fissures in Ni-based alloys probably arising from fission product tellurium attack. This paper summarizes results of corrosion tests conducted recently to study effect of oxidation state in selected fuel salt on tellurium attack and to develop means of controlling tellurium cracking in the special Ni-based alloys recently developed for molten salt actinide recycler and tranforming (MOSART) system. Tellurium corrosion of Ni-based alloys was tested at temperatures up to 750 °C in stressed and unloaded conditions in molten LiF-BeF2 salt mixture fueled by about 20 mol% of ThF4 and 2 mol% of UF4 at different [U(IV)]/[U(III)] ratios: 0.7, 4, 20, 100 and 500. Following Ni-based alloys (in mass%): HN80М-VI (Mo—12, Cr—7.6, Nb—1.5), HN80МТY (Mo—13, Cr—6.8, Al—1.1, Ti—0.9), HN80МТW (Mo—9.4, Cr—7.0, Ti—1.7, W—5.5) and ЕМ-721 (W—25.2, Cr—5.7, Ti—0.17) were used for the study in the corrosion facility. If the redox state the fuel salt is characterized by uranium ratio [U(IV)]/[U(III)] < 1 the alloys' specimens get a more negative stationary electrode potential than equilibrium electrode potentials of some uranium intermetallic compounds and alloys with nickel and molybdenum. This leads to spontaneous behavior of alloy formation processes on the specimens' surface and further diffusion of uranium deep into the metallic phase. As consequence of this films of intermetallic compounds and alloys of nickel, molybdenum, tungsten with uranium are formed on the alloys specimens' surface, and intergranular corrosion does not take place. In the fuel salt with [U(IV)]/[U(III)] = 4-20 the potentials of uranium alloy formation with the main components of the tested alloys are not reached, that's why alloys and intermetallic compounds are not formed on the surface of the investigated chromium-nickel alloys. Under such conditions any intergranular tellurium corrosion of the selected alloys does not occur. In the fuel salt with [U(IV)/]/[U(III)] = 100 the potentials of uranium alloy formation with the main components of the tested alloys are not also reached. Under such redox conditions any traces intergranular tellurium IGC on the HN80MTY and H80M-VI alloys specimens are not found. Certain signs of incipient IGC in the form of tellurium presence on the grain boundaries in the HN80MTB and EM-721 alloys surface layer and formation of not too deep cracks on HN80MTB alloy surface were revealed at [U(IV)/]/[U(III)] = 100. With this uranium ratio in the presence of corrosion products on the surface of all of the alloys films, containing tellurium, metals of the construction alloys and carbon, are formed. In the melt with [U(IV)]/[U(III)] = 500 in all of the alloys tested the tellurium IGC took place. The HN80MTY alloy shows the maximum resistance to tellurium IGC. The intensity of tellurium IGC of the alloy (the K parameter) is by 3-5 times lower as compared to other alloys. The EM-721 alloy has the minimal resistance to tellurium IGC (K = 9200 pc m/cm, the depth of cracks is up to 434 μm). The studies have shown, that the intensity of the nickel alloys IGC is controlled by the [U(IV)]/[U(III)] ratio, and its dependence on this parameter is of threshold character. Providing the uranium ratio value's monitoring and regulation, it is possible to control the tellurium corrosion and in such a way to eliminate IGC completely or to minimize its value. The alloys strength characteristics and their structure were changed insignificantly after testing within the [U(IV)]/[U(III)] range from 0.7 tо 100. The changes are not linked with the influence of fuel salt, containing tellurium additions, but are stipulated by alloys structure, temperature factor, exposure time and mechanical loads. Significant effect of tellurium cracking on the alloys (excepting HN80MTY) strength characteristics was established after corrosion testing with [U(IV)]/[U(III)] = 500. In the absence of IGC all of the alloys investigated have a good ductility at high strength characteristics. The disrupture of specimens under mechanical tests both before and after corrosion tests of all alloys except for ЕМ-721 proceeds on a ductile mechanism. On the EM-721 alloy specimens, both in their initial state and after corrosion testing, clear signs of brittle destruction, caused by heterogeneity of its structure due to the presence of tungsten phase, are very clearly observed. The presence of such phases increases the alloy IGC and leads to reduction of the alloy resistance tellurium damage. The HN80MTY alloy has the best corrosion and mechanical properties. It does not undergo tellurium IGC in the molten 75LiF-5BeF2-20ThF4 salt mixture fueled by about 2 mol% of UF4 with [U(IV)]/[U(III)] ratio ⩽ 100. The alloy has high resistance to tellurium cracking at [U(IV)]/[U(III)] = 500. The alloy can be recommended as the main construction material for the fuel circuit with selected salt composition up to temperature 750 °С.

Shield materials recommended for space power nuclear reactors

NASA Technical Reports Server (NTRS)

Kaszubinski, L. J.

1973-01-01

Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

Special nuclear material simulation device

DOEpatents

Leckey, John H.; DeMint, Amy; Gooch, Jack; Hawk, Todd; Pickett, Chris A.; Blessinger, Chris; York, Robbie L.

2014-08-12

An apparatus for simulating special nuclear material is provided. The apparatus typically contains a small quantity of special nuclear material (SNM) in a configuration that simulates a much larger quantity of SNM. Generally the apparatus includes a spherical shell that is formed from an alloy containing a small quantity of highly enriched uranium. Also typically provided is a core of depleted uranium. A spacer, typically aluminum, may be used to separate the depleted uranium from the shell of uranium alloy. A cladding, typically made of titanium, is provided to seal the source. Methods are provided to simulate SNM for testing radiation monitoring portals. Typically the methods use at least one primary SNM spectral line and exclude at least one secondary SNM spectral line.

Thermodynamic properties of La-Ga-Al and U-Ga-Al alloys and the separation factor of U/La couple in the molten salt-liquid metal system

NASA Astrophysics Data System (ADS)

Novoselova, A.; Smolenski, V.; Volkovich, V. A.; Ivanov, A. B.; Osipenko, A.; Griffiths, T. R.

2015-11-01

The electrochemical behaviour of lanthanum and uranium was studied in fused 3LiCl-2KCl eutectic and Ga-Al eutectic liquid metal alloy between 723 and 823 K. Electrode potentials were recorded vs. Cl-/Cl2 reference electrode and the temperature dependencies of the apparent standard potentials of La-(Ga-Al) and U-(Ga-Al) alloys were determined. Lanthanum and uranium activity coefficients and U/La couple separation factor were calculated. Partial excess free Gibbs energy, partial enthalpy of mixing and partial excess entropy of La-(Ga-Al) and U-(Ga-Al) alloys were estimated.

NUCLEAR FUEL COMPOSITION

DOEpatents

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

1960-05-31

A novel reactor composition for use in a self-sustaining fast nuclear reactor is described. More particularly, a fuel alloy comprising thorium and uranium-235 is de scribed, the uranium-235 existing in approximately the same amount that it is found in natural uranium, i.e., 1.4%.

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.

PYROMETALLURGICAL METHOD

DOEpatents

Nelson, P.A.

1961-07-18

The liquid--liquid extraction of plutonium by magnesium from uranium or uranium--chromium alloy is described. Calcium is added to magnesium in about eutectic proportions, which results in a purer plutonium.

TUNGSTEN INTERFERENCE IN VOLUMETRIC ANALYSIS OF URANIUM

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

Dufour, R.F.; Articolo, O.

1958-08-01

Tungsten was found to have a negligible effect on the determination of uranium in uranium-zirconium alloys by the Jones reductor-dichromate method used at KAPL. The tungstate ion interferred seriously and gave high results. However, the soluble tungsten was precipitated by intensive fuming with sulfuric acid and rendered ineffective in tbe subsequent oxidationreduction reactions of the uranium. (auth)

Creep behavior of uranium carbide-based alloys

NASA Technical Reports Server (NTRS)

Seltzer, M. S.; Wright, T. R.; Moak, D. P.

1975-01-01

The present work gives the results of experiments on the influence of zirconium carbide and tungsten on the creep properties of uranium carbide. The creep behavior of high-density UC samples follows the classical time-dependence pattern of (1) an instantaneous deformation, (2) a primary creep region, and (3) a period of steady-state creep. Creep rates for unalloyed UC-1.01 and UC-1.05 are several orders of magnitude greater than those measured for carbide alloys containing a Zr-C and/or W dispersoid. The difference in creep strength between alloyed and unalloyed materials varies with temperature and applied stress.

CAN HANDLING FIXTURES

DOEpatents

Kelman, Ler.R.; Yaggee, F.L.

1958-08-01

A sleeveless cauning apparatus is described for bonding and canning uranium fuel elements under the surface of a liquid bonding alloy. The can is supported on a pedestal by vertical pegs, and an adjustable collar is placed around the upper, open end of the can, which preferably is flared to assure accurate centering in the fixture and to guide the uranium slug into the can. The fixture with a can in place is then immersed in a liquid aluminum-silicon alloy and the can becomes filled with the liquid alloy. The slug is inserted by a slug guide located vertically above the can opening. The slug settles by gravity into the can, after which a cap is emplaced. A quenching tool lifts the capped can out of the bath by means of a slot provided for it in the pedestal. This apparatus provides a simple means of canning the slug without danger of injury to the uranium metal or the aluminum can.

THE ACCURATE DETERMINATION OF MICROGRAM AMOUNTS OF BORON IN ALUMINUM AND ALUMINUM-URANIUM ALLOYS BY THE METHYL BORATE-CURCUMIN-OXALIC ACID METHOD

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

Crocker, I.H.

1958-10-01

A method was developed for the deternninntion of boron in aluminum and aluminum--uranium alloys in which the boron concentration is 30 ppm or more. Boron is separated by distillation as methyl borate from a hydrochloric acid solution of the alloy and is determined spectrophotometrically by the boric acid-- curcumin-oxalic acid color reaction. A precision of plus or minus 2% is attain able when the determination is penformed with the utmost care. The accuracy is such that no bias need be given when a calibration curve is used. (auth)

HEAT TREATED U-Nb ALLOYS

DOEpatents

McGeary, R.K.; Justusson, W.M.

1959-11-24

A fuel element for a nuclear reactor is described comprising an alloy containing uranium and from 7 to 20 wt.% niobium, the alloy being substantially in the gamma phase and having been produced by working an ingot of the alloy into the desired shape, homogenizing it by annealing it at a temperature in the gamma phase field, and quenching it to retain the gamma phase structure of the alloy.

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

DOEpatents

Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

1999-01-01

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

THE ANALYSIS OF URANIUM-ZIRCONIUM ALLOYS

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

Milner, G.W.C.; Skewies, A.F.

1953-03-01

A satisfactory procedure is described for the analysis of uranium-zirconium alloys containing up to 25% zirconium. It is based on the separation of the zirconium from the uranium by dissolving the cupferron complex of the former element into chloroform. After the evaporation of the solvent from the combined organic extracts, the residue is ignited to zirconium oxide. The latter is then re-dissolved and zirconium is separated from other elements co-extracted in the solvent extraction procedure by precipitation with mandelic acid. The zirconium mandelate is finally ignited to oxide at 960 deg C. The uranium is separated from the aqueous solutionmore » remaining from the cupferron extraction by precipitating with tannin at a pH of 8; the precipitate being removed by filtration and then ignited a t 800 deg C. The residue is dissolved in nitric acid and the uranium is finally determined by precipitating as ammonium diuranate and then igniting to U{sub 3}O{sub 8}. (auth)« less

Investigation of americium-241 metal alloys for target applications. [Alloys with cerium

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

Conner, W.V.

1980-01-01

Several americium-241 metal alloys have been investigated for possible use in the Lawrence Livermore National Laboratory Radiochemical Diagnostic Tracer Program. Alloys investigated have included uranium-americium, aluminum-americium, and cerium-americium. Uranium-americium alloys with the desired properties proved to be difficult to prepare, and work with this alloy was discontinued. Aluminum-americium alloys were much easier to prepare, but the alloy consisted of an aluminum-americium intermetallic compound (AmAl/sub 4/) in an aluminum matrix. This alloy could be cast and formed into shapes, but the low density of aluminum, and other problems; made the alloy unsuitable for the intended application. Americium metal was found tomore » have a high solid solubility in cerium and alloys prepared from these two elements exhibited all of the properties desired for the tracer program application. Cerium-americium alloys containing up to 34 wt % americium have been prepared using both comelting and coreduction techniques. The latter technique involves coreduction of Ce F/sub 4/ and AmF/sub 4/ with calcium metal in a sealed reduction vessel. Casting techniques have been developed for preparing up to eight 0.87 inch (2.2 cm) diameter disks in a single casting, and cerium-americium metal alloy disks containing from 10 to 25 wt % americium-241 have been prepared using these techniques.« less

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

DOEpatents

Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

1999-03-23

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

METHOD OF SUPPRESSING UAl$sub 4$ FORMATION IN U-Al ALLOYS

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1960-08-23

A method is given for suppressing the formation of UAl/sub 4/ in uranium- - aluminum alloys, thereby rendering these alloys more easily workable. The method comprises incorporating in the base alloy a Group Four element selected from the group consisting of Si, Ti, Ge, Zr, and Sn, the addition preferably being within the range of 0.5to20at.%.

U-Zr alloy: XPS and TEM study of surface passivation

NASA Astrophysics Data System (ADS)

Paukov, M.; Tkach, I.; Huber, F.; Gouder, T.; Cieslar, M.; Drozdenko, D.; Minarik, P.; Havela, L.

2018-05-01

Surface reactivity of Uranium metal is an important factor limiting its practical applications. Bcc alloys of U with various transition metals are much less reactive than pure Uranium. So as to specify the mechanism of surface protection, we have been studying the U-20 at.% Zr alloy by photoelectron spectroscopy and transmission electron microscopy. The surface was studied in as-obtained state, in various stages of surface cleaning, and during an isochronal annealing cycle. The analysis based on U-4f, Zr-3p, and O-1 s spectra shows that a Zr-rich phase segregates at the surface at temperatures exceeding 550 K, which provides a self-assembled coating. The comparison of oxygen exposure of the stoichiometric and coated surfaces shows that the coating is efficiently preventing the oxidation of uranium even at elevated temperatures. The coating can be associated with the UZr2+x phase. TEM study indicated that the coating is about 20 nm thick. For the clean state, the U-4f core-level lines of the bcc alloy are practically identical to those of α-U, revealing similar delocalization of the 5f electronic states.

Magnesium transport extraction of transuranium elements from LWR fuel

DOEpatents

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

1992-01-01

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

PROTECTIVELY COVERED ARTICLE AND METHOD OF MANUFACTURE

DOEpatents

Plott, R.F.

1958-10-28

A method of casting a protective jacket about a ura nium fuel element that will bond completely to the uranium without the use of stringers or supports that would ordinarily produce gaps in the cast metal coating and bond is presented. Preformed endcaps of alumlnum alloyed with 13% silicon are placed on the ends of the uranium fuel element. These caps will support the fuel element when placed in a mold. The mold is kept at a ing alloy but below that of uranium so the cast metal jacket will fuse with the endcaps forming a complete covering and bond to the fuel element, which would otherwise oxidize at the gaps or discontinuities lefi in the coating by previous casting methods.

HEAT TREATED U-Mo ALLOY

DOEpatents

McGeary, R.K.; Justusson, W.M.

1960-02-23

A reactor fuel element comprising a gamma-phase alloy consisting of 11 to 16 wt.% of molyhdenum and the balance uranium, annealed between 350 and 525 deg C and quenched to preserve the gamma phase, is reported.

PREPARATION OF ACTINIDE-ALUMINUM ALLOYS

DOEpatents

Moore, R.H.

1962-09-01

BS>A process is given for preparing alloys of aluminum with plutonium, uranium, and/or thorium by chlorinating actinide oxide dissolved in molten alkali metal chloride with hydrochloric acid, chlorine, and/or phosgene, adding aluminum metal, and passing air and/or water vapor through the mass. Actinide metal is formed and alloyed with the aluminum. After cooling to solidification, the alloy is separated from the salt. (AEC)

DUCTILE URANIUM FUEL FOR NUCLEAR REACTORS AND METHOD OF MAKING

DOEpatents

Zegler, S.T.

1963-11-01

The fabrication process for a ductile nuclear fuel alloy consisting of uranium, fissium, and from 0.25 to 1.0 wt% of silicon or aluminum or from 0.25 to 2 wt% of titanium or yttrium is presented. (AEC)

Hot rolling of thick uranium molybdenum alloys

DOEpatents

DeMint, Amy L.; Gooch, Jack G.

2015-11-17

Disclosed herein are processes for hot rolling billets of uranium that have been alloyed with about ten weight percent molybdenum to produce cold-rollable sheets that are about one hundred mils thick. In certain embodiments, the billets have a thickness of about 7/8 inch or greater. Disclosed processes typically involve a rolling schedule that includes a light rolling pass and at least one medium rolling pass. Processes may also include reheating the rolling stock and using one or more heavy rolling passes, and may include an annealing step.

METAL COATING BATHS

DOEpatents

Robinson, J.W.

1958-08-26

A method is presented for restoring the effectiveness of bronze coating baths used for hot dip coating of uranium. Such baths, containing a high proportion of copper, lose their ability to wet uranium surfaces after a period of use. The ability of such a bath to wet uranium can be restored by adding a small amount of metallic aluminum to the bath, and skimming the resultant hard alloy from the surface.

Atomistic modeling of high temperature uranium-zirconium alloy structure and thermodynamics

NASA Astrophysics Data System (ADS)

Moore, A. P.; Beeler, B.; Deo, C.; Baskes, M. I.; Okuniewski, M. A.

2015-12-01

A semi-empirical Modified Embedded Atom Method (MEAM) potential is developed for application to the high temperature body-centered-cubic uranium-zirconium alloy (γ-U-Zr) phase and employed with molecular dynamics (MD) simulations to investigate the high temperature thermo-physical properties of U-Zr alloys. Uranium-rich U-Zr alloys (e.g. U-10Zr) have been tested and qualified for use as metallic nuclear fuel in U.S. fast reactors such as the Integral Fast Reactor and the Experimental Breeder Reactors, and are a common sub-system of ternary metallic alloys like U-Pu-Zr and U-Zr-Nb. The potential was constructed to ensure that basic properties (e.g., elastic constants, bulk modulus, and formation energies) were in agreement with first principles calculations and experimental results. After which, slight adjustments were made to the potential to fit the known thermal properties and thermodynamics of the system. The potentials successfully reproduce the experimental melting point, enthalpy of fusion, volume change upon melting, thermal expansion, and the heat capacity of pure U and Zr. Simulations of the U-Zr system are found to be in good agreement with experimental thermal expansion values, Vegard's law for the lattice constants, and the experimental enthalpy of mixing. This is the first simulation to reproduce the experimental thermodynamics of the high temperature γ-U-Zr metallic alloy system. The MEAM potential is then used to explore thermodynamics properties of the high temperature U-Zr system including the constant volume heat capacity, isothermal compressibility, adiabatic index, and the Grüneisen parameters.

Study of fluoride corrosion of nickel alloys

NASA Technical Reports Server (NTRS)

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

1969-01-01

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

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

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

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

1997-08-01

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

Coated Metal Articles and Method of Making

DOEpatents

Boller, Ernest R.; Eubank, Lowell D.

2004-07-06

The method of protectively coating metallic uranium which comprises dipping the metallic uranium in a molten alloy comprising about 20-75% of copper and about 80-25% of tin, dipping the coated uranium promptly into molten tin, withdrawing it from the molten tin and removing excess molten metal, thereupon dipping it into a molten metal bath comprising aluminum until it is coated with this metal, then promptly withdrawing it from the bath.

FUEL ELEMENTS AND METHOD OF MAKING

DOEpatents

Noland, R.A.; Marzano, C.

1958-08-19

A process is described of surface-impregnating bodies of metallic uranium with silicon. Silicon metal is added to or admixed with alkali metal selected from the group consisting of sodiunn, potassium, and sodiunnpotassium alloy. The uraniunn body is then immersed in the mixture obtained and the temperature is raised to between 425 and 600 deg C. The silicon is dissolved and deposits as a uranium-silicon compound on the uranium body.

Influence of a doping by Al stainless steel on kinetics and character of interaction with the metallic nuclear fuel

NASA Astrophysics Data System (ADS)

Nikitin, S. N.; Shornikov, D. P.; Tarasov, B. A.; Baranov, V. G.

2016-04-01

Metallic nuclear fuel is a perspective kind of fuel for fast reactors. In this paper we conducted a study of the interaction between uranium-molybdenum alloy and ferritic- martensitic steels with additions of aluminum at a temperature of 700 ° C for 25 hours. The rate constants of the interaction layer growth at 700 °C is about 2.8.10-14 m2/s. It is established that doping Al stainless steel leads to decrease in interaction with uranium-molybdenum alloys. The phase composition of the interaction layer is determined.

PLUTONIUM-URANIUM ALLOY

DOEpatents

Coffinberry, A.S.; Schonfeld, F.W.

1959-09-01

Pu-U-Fe and Pu-U-Co alloys suitable for use as fuel elements tn fast breeder reactors are described. The advantages of these alloys are ease of fabrication without microcracks, good corrosion restatance, and good resistance to radiation damage. These advantages are secured by limitation of the zeta phase of plutonium in favor of a tetragonal crystal structure of the U/sub 6/Mn type.

DIMENSIONALLY STABLE, CORROSION RESISTANT NUCLEAR FUEL

DOEpatents

Kittel, J.H.

1963-10-31

A method of making a uranium alloy of improved corrosion resistance and dimensional stability is described. The alloy contains from 0-9 weight per cent of an additive of zirconium and niobium in the proportions by weight of 5 to 1 1/ 2. The alloy is cold rolled, heated to two different temperatures, air-cooled, heated to a third temperature, and quenched in water. (AEC)

NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

DOEpatents

Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

1957-11-12

This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL

DOEpatents

Buyers, A.G.

1959-06-30

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

Powder formation of {gamma} uranium-molybdenum alloys via hydration-dehydration

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

Vaz de Oliveira, Fabio Branco; Durazzo, Michelangelo; Fontenele Urano de Carvalho, Elita

2008-07-15

Gamma uranium-molybdenum alloys has been considered as fuel phase in plate type fuel elements for MTR reactors, mainly due to their acceptable performance under irradiation and metallurgical processing. To its use as a dispersion phase in aluminum matrix, a necessary step is the conversion of the as cast structure into powder, and one of the techniques considered at IPEN / CNEN - Brazil is HDH (hydration-dehydration). The alloys were produced by the induction melting technique, and samples were obtained from the alloys for the thermal treatments, under constant flow of hydrogen, for temperatures varying from 400 deg C to 600more » deg C and times from 1 to 4 hours, followed by dehydration. A preliminary characterization of the powders was made and the curves of mass variation versus time were obtained and related to the powder characteristics. This paper describes the first results on the development of the technology to the powder formation of the (5 to 10) % weight molybdenum {gamma}-UMo alloys, and discusses some of its aspects, mainly those related to the {gamma} {yields} {alpha} equilibrium data. (author)« less

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

PubMed

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

2014-07-01

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

THE CHEMICAL ANALYSIS OF TERNARY ALLOYS OF PLUTONIUM WITH MOLYBDENUM AND URANIUM

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

Phillips, G.; Woodhead, J.; Jenkins, E.N.

1958-09-01

It is shown that the absorptiometric determination of molybdenum as thiocyanate may be used in the presence of plutonium. Molybdenum interferes with previously published methods for determining uranium and plutonium but conditlons have been established for its complete removal by solvent extraction of the compound with alpha -benzoin oxime. The previous methods for uranium and plutonium are satisfactory when applied to the residual aqueous phase following this solvent extraction. (auth)

Pumped lithium loop test to evaluate advanced refractory metal alloys and simulated nuclear fuel elements

NASA Technical Reports Server (NTRS)

Brandenburf, G. P.; Hoffman, E. E.; Smith, J. P.

1974-01-01

The performance was determined of refractory metal alloys and uranium nitride fuel element specimens in flowing 1900F (1083C) lithium. The results demonstrate the suitability of the selected materials to perform satisfactorily from a chemical compatibility standpoint.

Modeling Early-Stage Processes of U-10 Wt.%Mo Alloy Using Integrated Computational Materials Engineering Concepts

NASA Astrophysics Data System (ADS)

Wang, Xiaowo; Xu, Zhijie; Soulami, Ayoub; Hu, Xiaohua; Lavender, Curt; Joshi, Vineet

2017-12-01

Low-enriched uranium alloyed with 10 wt.% molybdenum (U-10Mo) has been identified as a promising alternative to high-enriched uranium. Manufacturing U-10Mo alloy involves multiple complex thermomechanical processes that pose challenges for computational modeling. This paper describes the application of integrated computational materials engineering (ICME) concepts to integrate three individual modeling components, viz. homogenization, microstructure-based finite element method for hot rolling, and carbide particle distribution, to simulate the early-stage processes of U-10Mo alloy manufacture. The resulting integrated model enables information to be passed between different model components and leads to improved understanding of the evolution of the microstructure. This ICME approach is then used to predict the variation in the thickness of the Zircaloy-2 barrier as a function of the degree of homogenization and to analyze the carbide distribution, which can affect the recrystallization, hardness, and fracture properties of U-10Mo in subsequent processes.

Strength and fracture of uranium, plutonium and several their alloys under shock wave loading

NASA Astrophysics Data System (ADS)

Golubev, V. K.

2012-08-01

Results on studying the spall fracture of uranium, plutonium and several their alloys under shock wave loading are presented in the paper. The problems of influence of initial temperature in a range of - 196 - 800∘C and loading time on the spall strength and failure character of uranium and two its alloys with molybdenum and both molybdenum and zirconium were studied. The results for plutonium and its alloy with gallium were obtained at a normal temperature and in a temperature range of 40-315∘C, respectively. The majority of tests were conducted with the samples in the form of disks 4 mm in thickness. They were loaded by the impact of aluminum plates 4 mm thick through a copper screen 12 mm thick serving as the cover or bottom part of a special container. The character of spall failure of materials and the damage degree of samples were observed on the longitudinal metallographic sections of recovered samples. For a concrete test temperature, the impact velocity was sequentially changed and therefore the loading conditions corresponding to the consecutive transition from microdamage nucleation up to complete macroscopic spall fracture were determined. The conditions of shock wave loading were calculated using an elastic-plastic computer program. The comparison of obtained results with the data of other researchers on the spall fracture of examined materials was conducted.

High density dispersion fuel

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

Hofman, G.L.

1996-09-01

A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm{sup 3} of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm{sup {minus}3} with U{sub 3}Si{sub 2} as fuel. High-density uranium compounds offer no real density advantage over U{sub 3}Si{sub 2} and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U{sub 3}Si has approximately a 30% higher uranium density but the density of the U{sub 6}X compounds would yield the factormore » 1.5 needed to achieve 9 g cm{sup {minus}3} uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure {alpha}-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic {gamma} phase at low temperatures where normally {alpha} phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing.« less

Cubic γ-phase U-Mo alloys synthesized by splat-cooling

NASA Astrophysics Data System (ADS)

Kim-Ngan, Nhu-T. H.; Tkach, I.; Mašková, S.; Havela, L.; Warren, A.; Scott, T.

2013-09-01

U-Mo alloys are the most promising materials fulfilling the requirements of using low enriched uranium (LEU) fuel in research reactors. From a fundamental standpoint, it is of interest to determine the basic thermodynamic properties of the cubic γ-phase U-Mo alloys. We focus our attention on the use of Mo doping together with ultrafast cooling (with high cooling rates ⩾106 K s-1), which helps to maintain the cubic γ-phase in U-Mo system to low temperatures and on determination of the low-temperature properties of these γ-U alloys. Using a splat cooling method it has been possible to maintain some fraction of the high-temperature γ-phase at room temperature in pure uranium. U-13 at.% Mo splat clearly exhibits the pure γ-phase structure. All the splats become superconducting with Tc in the range from 1.24 K (pure U splat) to 2.11 K (U-15 at.% Mo). The γ-phase in U-Mo alloys undergoes eutectoid decomposition to form equilibrium phases of orthorhombic α-uranium and tetragonal γ‧-phase upon annealing at 500 °C, while annealing at 800 °C has stabilized the initial γ phase. The α-U easily absorbs a large amount of hydrogen (UH3 hydride), while the cubic bcc phase does not absorb any detectable amount of hydrogen at pressures below 1 bar and at room temperature. At 80 bar, the U-15 at.% Mo splat becomes powder consisting of elongated particles of 1-2 mm, revealing amorphous state.

O-Pu-U (Oxygen-Plutonium-Uranium)

NASA Astrophysics Data System (ADS)

Materials Science International Team MSIT

This document is part of Subvolume C4 'Non-Ferrous Metal Systems. Part 4: Selected Nuclear Materials and Engineering Systems' of Volume 11 'Ternary Alloy Systems - Phase Diagrams, Crystallographic and Thermodynamic Data critically evaluated by MSIT®' of Landolt-Börnstein - Group IV 'Physical Chemistry'. It provides data of the ternary system Oxygen-Plutonium-Uranium.

Cobalt-Free Permanent Magnet Alloys.

DTIC Science & Technology

1984-10-01

carbide co- UC CbC lumbium carbide M003 Uranium carbide - tho- UC 2 25ThC rium carbide ZrO2 MgO WOs Use of this Process for MnAlC As indicated in the...cobalt. Free World Cobal Consumption Estimated Breakdown by End Uses Magnetic alloys 20% Cemented carbides - 5% 30 SuPerolloy _ 15% Other steels and...would normally result in the formation of binary alloy of TbFe 2 and preventing the formation of amorphous alloy (Fe-B) contain- ing Tb. The

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.

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.

FORMING TUBES AND RODS OF URANIUM METAL BY EXTRUSION

DOEpatents

Creutz, E.C.

1959-01-27

A method and apparatus are presented for the extrusion of uranium metal. Since uranium is very brittle if worked in the beta phase, it is desirable to extrude it in the gamma phase. However, in the gamma temperature range thc uranium will alloy with the metal of the extrusion dic, and is readily oxidized to a great degree. According to this patent, uranium extrusion in thc ganmma phase may be safely carried out by preheating a billet of uranium in an inert atmosphere to a trmperature between 780 C and 1100 C. The heated billet is then placed in an extrusion apparatus having dies which have been maintained at an elevated temperature for a sufficient length of time to produce an oxide film, and placing a copper disc between the uranium billet and the die.

FUEL ELEMENT AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.

1961-04-25

A nuclear fuel element in the form of a wire is reported. A bar of uranium is enclosed in a thin layer of aluminum and the composite is sheathed in beryllium, zirconium, or stainnless steel. The sheathed article is then drawn to wire form, heated to alloy the aluminum with both uranium and sheath, and finally cold worked.

Atomistic simulation of cubic and tetragonal phases of U-Mo alloy: Structure and thermodynamic properties

NASA Astrophysics Data System (ADS)

Starikov, S. V.; Kolotova, L. N.; Kuksin, A. Yu.; Smirnova, D. E.; Tseplyaev, V. I.

2018-02-01

We studied structure and thermodynamic properties of cubic and tetragonal phases of pure uranium and U-Mo alloys using atomistic simulations: molecular dynamics and density functional theory. The main attention was paid to the metastable γ0 -phase that is formed in U-Mo alloys at low temperature. Structure of γ0 -phase is similar to body-centered tetragonal (bct) lattice with displacement of a central atom in the basic cell along [ 001 ] direction. Such displacements have opposite orientations for part of the neighbouring basic cells. In this case, such ordering of the displacements can be designated as antiferro-displacement. Formation of such complex structure may be interpreted through forming of short U-U bonds. At heating, the tetragonal structure transforms into cubic γs -phase, still showing ordering of central atom displacements. With rise in temperature, γs -phase transforms to γ-phase with a quasi body-centered cubic (q-bcc) lattice. The local positions of uranium atoms in γ-phase correspond to γs -phase, however, orientations of the central atom displacements become disordered. Transition from γ0 to γ can be considered as antiferro-to paraelastic transition of order-disorder type. This approach to the structure description of uranium alloy allows to explain a number of unusual features found in the experiments: anisotropy of lattice at low temperature; remarkably high self-diffusion mobility in γ-phase; decreasing of electrical resistivity at heating for some alloys. In addition, important part of this work is the development of new interatomic potential for U-Mo system made with taking into account details of studied structures.

Fabrication of thorium bearing carbide fuels

DOEpatents

Gutierrez, Rueben L.; Herbst, Richard J.; Johnson, Karl W. R.

1981-01-01

Thorium-uranium carbide and thorium-plutonium carbide fuel pellets have been fabricated by the carbothermic reduction process. Temperatures of 1750.degree. C. and 2000.degree. C. were used during the reduction cycle. Sintering temperatures of 1800.degree. C. and 2000.degree. C. were used to prepare fuel pellet densities of 87% and >94% of theoretical, respectively. The process allows the fabrication of kilogram quantities of fuel with good reproducibility of chemicals and phase composition. Methods employing liquid techniques that form carbide microspheres or alloying-techniques which form alloys of thorium-uranium or thorium-plutonium suffer from limitation on the quantities processed of because of criticality concerns and lack of precise control of process conditions, respectively.

METHOD OF MAKING ALLOYS OF BERYLLIUM WITH PLUTONIUM AND THE LIKE

DOEpatents

Runnals, O.J.C.

1959-02-24

The production of alloys of beryllium with one or more of the metals uranium, plutonium, actinium, americium, curium, thorium, and cerium are described. A halide salt of the metal to be alloyed with the beryllium is heated at 1300 deg C in the presence of beryllium to reduce the halide to metal and cause the latter to alloy directly with the beryllium. Although the heavy metal halides are more stable, thermodynamically, than the beryllium halides, the reducing reaction proceeds to completion if the beryllium halide product is continuously removed by vacuum distillation.

Apparatus to recover tritium from tritiated molecules

DOEpatents

Swansiger, William A.

1988-01-01

An apparatus for recovering tritium from tritiated compounds is provided, including a preheater for heating tritiated water and other co-injected tritiated compounds to temperatures of about 600.degree. C. and a reactor charged with a mixture of uranium and uranium dioxide for receiving the preheated mixture. 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. 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.

Compatibility of buffered uranium carbides with tungsten.

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1971-01-01

Results of compatibility tests between tungsten and hyperstoichiometric uranium carbide alloys run at 1800 C for 1000 and 2500 hours. These tests compared tungsten-buffered uranium carbide with tungsten-buffered uranium-zirconium carbide. The zirconium carbide addition appeared to widen the homogeneity range of the uranium carbide, making additional carbon available for reaction. Reaction layers could be formed by either of two diffusion paths, one producing UWC2, while the second resulted in the formation of W2C. UWC2 acts as a diffusion barrier for carbon and slows the growth of the reaction layer with time, while carbon diffusion is relatively rapid in W2C, allowing equilibrium to be reached in less than 2500 hours at a temperature of 1800 C.

Kr ion irradiation study of the depleted-uranium alloys

NASA Astrophysics Data System (ADS)

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

2010-12-01

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

Microstructure of RERTR DU-Alloys Irradiated with Krypton Ions

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

J. Gan; D. Keiser; D. Wachs

2009-11-01

Fuel development for reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium fuels that can be employed to replace existing high enrichment uranium fuels currently used in many research and test reactors worldwide. Radiation stability of the interaction product formed at fuel-matrix interface has a strong impact on fuel performance. Three depleted uranium alloys are cast that consist of the following 5 phases of interest to be investigated: U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, U6Mo4Al43 and UAl4. Irradiation of TEM disc samples with 500 keV Kr ions at 200?C to high doses up tomore » ~100 dpa were conducted using an intermediate voltage electron microscope equipped with an ion accelerator. The irradiated microstructure of the 5 phases is characterized using transmission electron microscopy. The results will be presented and the implication of the observed irradiated microstructure on the fuel performance will be discussed.« less

Corrosion-resistant uranium

DOEpatents

Hovis, V.M. Jr.; Pullen, W.C.; Kollie, T.G.; Bell, R.T.

1981-10-21

The present invention is directed to the protecting of uranium and uranium alloy articles from corrosion by providing the surfaces of the articles with a layer of an ion-plated metal selected from aluminum and zinc to a thickness of at least 60 microinches and then converting at least the outer surface of the ion-plated layer of aluminum or zinc to aluminum chromate or zinc chromate. This conversion of the aluminum or zinc to the chromate form considerably enhances the corrosion resistance of the ion plating so as to effectively protect the coated article from corrosion.

Corrosion-resistant uranium

DOEpatents

Hovis, Jr., Victor M.; Pullen, William C.; Kollie, Thomas G.; Bell, Richard T.

1983-01-01

The present invention is directed to the protecting of uranium and uranium alloy articles from corrosion by providing the surfaces of the articles with a layer of an ion-plated metal selected from aluminum and zinc to a thickness of at least 60 microinches and then converting at least the outer surface of the ion-plated layer of aluminum or zinc to aluminum chromate or zinc chromate. This conversion of the aluminum or zinc to the chromate form considerably enhances the corrosion resistance of the ion plating so as to effectively protect the coated article from corrosion.

China and Proliferation of Weapons of Mass Destruction and Missiles: Policy Issues

DTIC Science & Technology

2010-08-16

nuclear weapons facilities, while experts from China worked at a uranium mine at Saghand and a centrifuge facility (for uranium enrichment) near...brief interruptions.”85 84 Barbara Opall -Rome and Vago Muradian, “Bush Privately Lauds...confiscated a rare metal used to produce alloy steel (called vanadium) being smuggled to North Korea. In the same month, China’s NHI Shenyang Mining

Method of making alloys of beryllium with plutonium and the like

DOEpatents

Runnals, O J.C.

1959-02-24

The production or alloys of beryllium with one or more of the metals uranium, plutonium, actinium, americium, curium, thorium, and cerium is described. A halide salt or the metal to be alloyed with the beryllium is heated at l3O0 deg C in the presence of beryllium to reduce the halide to metal and cause the latter to alloy directly with the beryllium. Although the heavy metal halides are more stable, thermodynamically, than the beryllium halides, the reducing reaction proceeds to completion if the beryllium halide product is continuously removed by vacuum distillation.

Debye temperatures and magnetic structures of UFe xAl 12- x (3.6⩽ x⩽5) intermetallic alloys

NASA Astrophysics Data System (ADS)

Rećko, K.; Dobrzyński, L.; Szymański, K.; Hoser, A.

2000-03-01

Uranium ternary compounds UFe xAl 12- x crystallize in a body-centred tetragonal structure ThMn 12 (I 4/mmm No.139). The neutron powder diffraction, magnetization measurements as well as Mössbauer investigations clearly indicate the magnetic ordering within the iron sites. The rearrangement of iron magnetic moments from uncompensated antiferromagnetic system in UFe xAl 12- x with x<4, through coexistence of antiferro- and ferromagnetic iron components (4⩽ x<5) to pure ferromagnetic ordering for alloy with x=5 is observed. The neutron diffraction studies of magnetic structures of the aforementioned powder samples show a very rich world of possible uranium-iron magnetic interactions. For all these alloys the magnetic neutron scattering is generally weak in comparison to the nuclear one. Because of identical chemical and magnetic unit cells there are no pure magnetic reflections. Therefore, in order to extract magnetic part of the scattering one should be particularly careful in taking proper account of the thermal vibration effects.

Alternative Anodes for the Electrolytic Reduction of Uranium Dioxide

NASA Astrophysics Data System (ADS)

Merwin, Augustus

Reprocessing of spent nuclear fuel is an essential step in closing the nuclear fuel cycle. In order to consume current stockpiles, ceramic uranium dioxide spent nuclear fuel will be subjected to an electrolytic reduction process. The current reduction process employs a platinum anode and a stainless steel alloy 316 cathode in a molten salt bath consisting of LiCl-2wt% Li 2O and occurs at 700°C. A major shortcoming of the existing process is the degradation of the platinum anode under the severely oxidizing conditions encountered during electrolytic reduction. This work investigates alternative anode materials for the electrolytic reduction of uranium oxide. The high temperature and extreme oxidizing conditions encountered in these studies necessitated a unique set of design constraints on the system. Thus, a customized experimental apparatus was designed and constructed. The electrochemical experiments were performed in an electrochemical reactor placed inside a furnace. This entire setup was housed inside a glove box, in order to maintain an inert atmosphere. This study investigates alternative anode materials through accelerated corrosion testing. Surface morphology was studied using scanning electron microscopy. Surface chemistry was characterized using energy dispersive spectroscopy and Raman spectroscopy. Electrochemical behavior of candidate materials was evaluated using potentiodynamic polarization characteristics. After narrowing the number of candidate electrode materials, ferrous stainless steel alloy 316, nickel based Inconel 718 and elemental tungsten were chosen for further investigation. Of these materials only tungsten was found to be sufficiently stable at the anodic potential required for electrolysis of uranium dioxide in molten salt. The tungsten anode and stainless steel alloy 316 cathode electrode system was studied at the required reduction potential for UO2 with varying lithium oxide concentrations. Electrochemical impedance spectroscopy showed mixed (kinetic and diffusion) control and an overall low impedance due to extreme corrosion. It was observed that tungsten is sufficiently stable in LiCl - 2wt% Li 2O at 700°C at the required anodic potential for the reduction of uranium oxide. This study identifies tungsten to be a superior anode material to platinum for the electrolytic reduction of uranium oxide, both in terms of superior corrosion behavior and reduced cost, and thus recommends that tungsten be further investigated as an alternative anode for the electrolytic reduction of uranium dioxide.

Salt transport extraction of transuranium elements from LWR fuel

DOEpatents

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

1992-11-03

A process is described for separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl[sub 2] and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750 C to about 850 C to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl[sub 2] having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO[sub 2]. The Ca metal and CaCl[sub 2] is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including MgCl[sub 2] to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy. 2 figs.

Salt transport extraction of transuranium elements from lwr fuel

DOEpatents

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

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750.degree. C. to about 850.degree. C. to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including Mg Cl.sub.2 to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy.

Spall fracture and strength of uranium, plutonium and their alloys under shock wave loading

NASA Astrophysics Data System (ADS)

Golubev, Vladimir

2015-06-01

Numerous results on studying the spall fracture phenomenon of uranium, two its alloys with molybdenum and zirconium, plutonium and its alloy with gallium under shock wave loading are presented in the paper. The majority of tests were conducted with the samples in the form of disks 4mm in thickness. They were loaded by the impact of aluminum plates 4mm thick through a copper screen serving as the cover or bottom part of a special container. The initial temperature of samples was changed in the range of -196 - 800 C degree for uranium and 40 - 315 C degree for plutonium. The character of spall failure of materials and the degree of damage for all tested samples were observed on the longitudinal metallographic sections of recovered samples. For a concrete test temperature, the impact velocity was sequentially changed and therefore the loading conditions corresponding to the consecutive transition from microdamage nucleation up to complete macroscopic spall fracture were determined. Numerical calculations of the conditions of shock wave loading and spall fracture of samples were performed in the elastoplastic approach. Several two- and three-dimensional effects of loading were taken into account. Some results obtained under conditions of intensive impulse irradiation and intensive explosive loading are presented too. The rather complete analysis and comparison of obtained results with the data of other researchers on the spall fracture of examined materials were conducted.

Interdiffusion, Intrinsic Diffusion, Atomic Mobility, and Vacancy Wind Effect in γ(bcc) Uranium-Molybdenum Alloy

NASA Astrophysics Data System (ADS)

Huang, Ke; Keiser, Dennis D.; Sohn, Yongho

2013-02-01

U-Mo alloys are being developed as low enrichment uranium fuels under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. In order to understand the fundamental diffusion behavior of this system, solid-to-solid pure U vs Mo diffusion couples were assembled and annealed at 923 K, 973 K, 1073 K, 1173 K, and 1273 K (650 °C, 700 °C, 800 °C, 900 °C, and 1000 °C) for various times. The interdiffusion microstructures and concentration profiles were examined via scanning electron microscopy and electron probe microanalysis, respectively. As the Mo concentration increased from 2 to 26 at. pct, the interdiffusion coefficient decreased, while the activation energy increased. A Kirkendall marker plane was clearly identified in each diffusion couple and utilized to determine intrinsic diffusion coefficients. Uranium intrinsically diffused 5-10 times faster than Mo. Molar excess Gibbs free energy of U-Mo alloy was applied to calculate the thermodynamic factor using ideal, regular, and subregular solution models. Based on the intrinsic diffusion coefficients and thermodynamic factors, Manning's formalism was used to calculate the tracer diffusion coefficients, atomic mobilities, and vacancy wind parameters of U and Mo at the marker composition. The tracer diffusion coefficients and atomic mobilities of U were about five times larger than those of Mo, and the vacancy wind effect increased the intrinsic flux of U by approximately 30 pct.

Computational and Experimental Studies of Microstructure-Scale Porosity in Metallic Fuels for Improved Gas Swelling Behavior

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

Mllett, Paul; McDeavitt, Sean; Deo, Chaitanya

This proposal will investigate the stability of bimodal pore size distributions in metallic uranium and uranium-zirconium alloys during sintering and re-sintering annealing treatments. The project will utilize both computational and experimental approaches. The computational approach includes both Molecular Dynamics simulations to determine the self-diffusion coefficients in pure U and U-Zr alloys in single crystals, grain boundaries, and free surfaces, as well as calculations of grain boundary and free surface interfacial energies. Phase-field simulations using MOOSE will be conducted to study pore and grain structure evolution in microstructures with bimodal pore size distributions. Experiments will also be performed to validate themore » simulations, and measure the time-dependent densification of bimodal porous compacts.« less

Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

2002-08-01

Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

Processing of U-2.5Zr-7.5Nb and U-3Zr-9Nb alloys by sintering process

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

Dos Santos, A. M. M.; Ferraz, W. B.; Lameiras, F. S.

2012-07-01

To minimize the risk of nuclear proliferation, there is worldwide interest in reducing fuel enrichment of research and test reactors. To achieve this objective while still guaranteeing criticality and cycle length requirements, there is need of developing high density uranium metallic fuels. Alloying elements such as Zr, Nb and Mo are added to uranium to improve fuel performance in reactors. In this context, the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) is developing the U-2.5Zr-7.5Nb and U-3Zr-9Nb (weight %) alloys by the innovative process of sintering that utilizes raw materials in the form of powders. The powders were pressed atmore » 400 MPa and then sintered under a vacuum of about 1x10{sup -4} Torr at temperatures ranging from 1050 deg. to 1500 deg.C. The densities of the alloys were measured geometrically and by hydrostatic method and the phases identified by X ray diffraction (XRD). The microstructures of the pellets were observed by scanning electron microscopy (SEM) and the alloying elements were analyzed by energy dispersive X-ray spectroscopy (EDS). The results obtained showed the fuel density to slightly increase with the sintering temperature. The highest density achieved was approximately 80% of theoretical density. It was observed in the pellets a superficial oxide layer formed during the sintering process. (authors)« less

Structural, microstructural and thermal analysis of U-(6-x)Zr-xNb alloys (x = 0, 2, 4, 6)

NASA Astrophysics Data System (ADS)

Kaity, Santu; Banerjee, Joydipta; Parida, S. C.; Bhasin, Vivek

2018-06-01

Uranium-rich U-Zr-Nb alloy is considered as a good alternative fuel for fast reactors from the perspective of excellent dimensional stability and desired thermo-physical properties to achieve higher burnup. Detailed investigations related to the structural and microstructural characterization, thermal expansion, phase transformation, microhardness were carried out on U-6Zr, U-4Zr-2Nb, U-2Zr-4Nb and U-6Nb alloys (composition in wt%) where the total amount of alloying elements was restricted to 6 wt%. Structural, microstructural and thermal analysis studies revealed that these alloys undergo a series of transformations from high temperature bcc γ-phase to a variety of equilibrium and intermediate phases depending upon alloy composition, cooling rate and quenching. The structural analysis was carried out by Rietveld refinement. The data of U-Nb and U-Zr-Nb alloys have been highlighted and compared with binary U-Zr alloy.

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

1993-09-01

The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

High temperature fuel/emitter system for advanced thermionic fuel elements

NASA Astrophysics Data System (ADS)

Moeller, Helen H.; Bremser, Albert H.; Gontar, Alexander; Fiviesky, Evgeny

1997-01-01

Specialists in space applications are currently focusing on bimodal power systems designed to provide both electric power and thermal propulsion (Kennedy, 1994 and Houts, 1995). Our work showed that thermionics is a viable technology for nuclear bimodal power systems. We demonstrated that materials for a thermionic fuel-emitter combination capable of performing at operating temperatures of 2473 K are not only possible but available. The objective of this work, funded by the US Department of Energy, Office of Space and Defense Power Systems, was to evaluate the compatibility of fuel material consisting of an uranium carbide/tantalum carbide solid solution with an emitter material consisting of a monocrystalline tungsten-niobium alloy. The uranium loading of the fuel material was 70 mole% uranium carbide. The program was successfully accomplished by a B&W/SIA LUTCH team. Its workscope was integrated with tasks being performed at both Babcock & Wilcox, Lynchburg Research Center, Lynchburg, Virginia, and SIA LUTCH, Podolsk, Russia. Samples were fabricated by LUTCH and seven thermal tests were performed in a hydrogen atmosphere. The first preliminary test was performed at 2273 K by LUTCH, and the remaining six tests were performed At B&W. Three tests were performed at 2273 K, two at 2373 K, and the final test at 2473 K. The results showed that the fuel and emitter materials were compatible in the presence of hydrogen. No evidence of liquid formation, dissolution of the uranium carbide from the uranium carbide/tantalum carbide solid solution, or diffusion of the uranium into the monocrystalline tungsten alloy was observed. Among the highlights of the program was the successful export of the fuel samples from Russia and their import into the US by commercial transport. This paper will discuss the technical aspects of this work.

4. DETAIL VIEW (SIDE A) OF HANDMADE STEEL BOX ASSOCIATED ...

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

4. DETAIL VIEW (SIDE A) OF HANDMADE STEEL BOX ASSOCIATED WITH THE DEPLETED URANIUM ALLOY DEVELOPMENT AND COMPONENT FABRICATION PROCESS. THE BOX WAS USED TO TRANSFER HEATED BLOCKS OF METAL (SHOWN IN THE OPENED DOOR) FROM THE MOLTEN SALT BATH TO THE ROLLER LINES. (4/28/62) - Rocky Flats Plant, Uranium Rolling & Forming Operations, Southeast section of plant, southeast quadrant of intersection of Central Avenue & Eighth Street, Golden, Jefferson County, CO

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.

NEUTRONIC REACTOR FUEL COMPOSITION

DOEpatents

Thurber, W.C.

1961-01-10

Uranium-aluminum alloys in which boron is homogeneously dispersed by adding it as a nickel boride are described. These compositions have particular utility as fuels for neutronic reactors, boron being present as a burnable poison.

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.

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

HEU Holdup Measurements in 321-M B and Spare U-Al Casting Furnaces

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

Salaymeh, S.R.

The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Decontamination Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. This report covers holdup measurements in two uranium aluminum alloy (U-Al) casting furnaces. Our results indicate an upper limit of 235U content for the B and Spare furnaces of 51 and 67 g respectively. This report discusses themore » methodology, non-destructive assay (NDA) measurements, and results of the uranium holdup on the two furnaces.« less

Use of ion beams to simulate reaction of reactor fuels with their cladding

NASA Astrophysics Data System (ADS)

Birtcher, R. C.; Baldo, P.

2006-01-01

Processes occurring within reactor cores are not amenable to direct experimental observation. Among major concerns are damage, fission gas accumulation and reaction between the fuel and its cladding all of which lead to swelling. These questions can be investigated through simulation with ion beams. As an example, we discuss the irradiation driven interaction of uranium-molybdenum alloys, intended for use as low-enrichment reactor fuels, with aluminum, which is used as fuel cladding. Uranium-molybdenum coated with a 100 nm thin film of aluminum was irradiated with 3 MeV Kr ions to simulate fission fragment damage. Mixing and diffusion of aluminum was followed as a function of irradiation with RBS and nuclear reaction analysis using the 27Al(p,γ)28Si reaction which occurs at a proton energy of 991.9 keV. During irradiation at 150 °C, aluminum diffused into the uranium alloy at a irradiation driven diffusion rate of 30 nm2/dpa. At a dose of 90 dpa, uranium diffusion into the aluminum layer resulted in formation of an aluminide phase at the initial interface. The thickness of this phase grew until it consumed the aluminum layer. The rapid diffusion of Al into these reactor fuels may offer explanation of the observation that porosity is not observed in the fuel particles but on their periphery.

Monte Carlo Criticality Analysis of Simple Geometrics COntaining Tungsten Rhenium Alloys Engrained with Uranium Dioxide and Uranium Mononitride

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

Jonathan A. Webb; Indrajit Charit

2011-08-01

The critical mass and dimensions of simple geometries containing highly enriched uraniumdioxide (UO2) and uraniummononitride (UN) encapsulated in tungsten-rhenium alloys are determined using MCNP5 criticality calculations. Spheres as well as cylinders with length to radius ratios of 1.82 are computationally built to consist of 60 vol.% fuel and 40 vol.% metal matrix. Within the geometries the uranium is enriched to 93 wt.% uranium-235 and the rhenium content within the metal alloy was modeled over a range of 0 to 30 at.%. The spheres containing UO2 were determined to have a critical radius of 18.29 cm to 19.11 cm and amore » critical mass ranging from 366 kg to 424 kg. The cylinders containing UO2 were found to have a critical radius ranging from 17.07 cm to 17.844 cm with a corresponding critical mass of 406 kg to 471 kg. Spheres engrained with UN were determined to have a critical radius ranging from 14.82 cm to 15.19 cm and a critical mass between 222 kg and 242 kg. Cylinders which were engrained with UN were determined to have a critical radius ranging from 13.811 cm to 14.155 cm with a corresponding critical mass of 245 kg to 267 kg. The critical geometries were also computationally submerged in a neutronaically infinite medium of fresh water to determine the effects of rhenium addition on criticality accidents due to water submersion. The monte carlo analysis demonstrated that rhenium addition of up to 30 at.% can reduce the excess reactivity due to water submersion by up to $5.07 for UO2 fueled cylinders, $3.87 for UO2 fueled spheres and approximately $3.00 for UN fueled spheres and cylinders.« less

NPF MECHANICAL CELL NaK DISPOSAL AND FUME ABATEMENT

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

Rey, G.

Some of the fuels originally scheduled for processing in the nonproduction fuel (NPF) processing program incorporated sodium or sodium- potassium alloy (NaK) as the bonding material between stainless-steel cladding and the uranium or uranium-molybdenum alloy core. Because of the special hazards involved in handling NaK, studies were made to determine safe methods for processing NaK-containing fuels. An underwater NaK dispensing system was installed, and tests were made to determine the characteristics of the NaK-water reaction. The equipment consisted of a dispenser, reaction pan, and off-gas scrubber. After initinl studies, a prototype test was made wherein U-Mo canned slugs containing NaKmore » reservoirs were hack sawed underwater. The studies demonstrated that the NaK reservoirs can be safely deactivated by hack sawing under a submerged hood in a shallow water bath. (W.L.H.)« less

Potential late health effects of depleted uranium and tungsten used in armor-piercing munitions: comparison of neoplastic transformation and genotoxicity with the known carcinogen nickel.

PubMed

Miller, Alexandra C; Xu, Jiaquan; Stewart, Michael; Prasanna, Pataje G S; Page, Natalie

2002-02-01

Limited data exist to permit an accurate assessment of risks for carcinogenesis and mutagenesis from embedded fragments or inhaled particulates of depleted uranium (DU). Ongoing studies have been designed to provide information about the carcinogenic potential of DU using in vitro and in vivo assessments of morphological transformation as well as cytogenetic, mutagenic, and oncogenic effects. For comparison, we also examined tungsten alloys used in military projectiles and the known carcinogen nickel. Quantitative and qualitative in vitro transformation studies were done to assess the carcinogenic potential of radiation and chemical hazards. Using a human osteosarcoma cell model, we demonstrated that soluble and insoluble DU compounds can transform cells to the tumorigenic phenotype, as characterized by morphological, biochemical, and oncogenic changes consistent with tumor cell behavior. Tungsten alloys and nickel were also shown to be neoplastic transforming agents, although at a frequency less than that of DU. Sister chromatid exchange, micronuclei, and alkaline filter elution assays showed DU and tungsten alloys were genotoxic. Exposure to a nontoxic, nontransforming dose of DU induced a small but statistically significant increase in the number of dicentrics formed in cells. These results suggest that long-term exposure to DU or tungsten alloys could be critical to the development of neoplastic disease in humans and that additional studies are needed.

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

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

Magoulas, V. E.

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium,more » and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.« less

Method for improving the mechanical properties of uranium-1 to 3 wt % zirconium alloy

DOEpatents

Anderson, R.C.

1983-11-22

A uranium-1 to 3 wt % zirconium alloy characterized by high strength, high ductility and stable microstructure is fabricated by an improved thermal mechanical process. A homogenous ingot of the alloy which has been reduced in thickness of at least 50% in the two-step forging operation, rolled into a plate with a 75% reduction and then heated in vacuum at a temperature of about 750 to 850/sup 0/C and then quenched in water, is subjected to further thermal-mechanical operation steps to increase the compressive yield strength approximately 30%, stabilize the microstructure, and decrease the variations in mechanical properties throughout the plate is provided. These thermal-mechanical steps are achieved by cold rolling the quenchd plate to reduce the thickness thereof about 8 to 12%, aging the cold rolled plate at a first temperature of about 325 to 375/sup 0/C for five to six hours and then aging the plate at a higher temperature ranging from 480 to 500/sup 0/C for five to six hours prior to cooling the billet to ambient conditions and sizing the billet or plate into articles provides the desired increase in mechanical properties and phase stability throughout the plate.

Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature

NASA Technical Reports Server (NTRS)

Rohal, R. G.; Tambling, T. N.; Smith, R. L.

1973-01-01

Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.

REGENERATION OF REACTOR FUEL ELEMENTS

DOEpatents

Lyon, W.L.

1960-04-01

A process is described for concentrating uranium and/or plutonium metal in aluminum alloys in which the actinide content was partially consumed by neutron bombardinent. Two embodiments are claimed: Either the alloy is heated, together with zinc chloride to at least 1000 deg C whereby some aluminum, in the form of aluminum chloride, and any zinc formed volatilize; or else aluminum fluoride is added and reacted at 800 to 1000 deg O and substmospheric pressure whereby pant of the aluminum volatilizes and aluminum subfluoride.

Plasma Processing Systems for the Manufacture of Refractory Metals and their Alloys for Military Needs

DTIC Science & Technology

1978-10-09

melting point is around 4000*K. An exceedingly interesting feature of these solidification composites is the formation of fibrous MC type carbide ...the matrix could be refractory metal binary alloys with copper or uranium and the eutectic phase could be carbide of tungsten, * molybdenum, tantalum or...42 Accs -n or - *DTTI Tf Avn ! -7ll ’ i CrDi t , l’’*i,;. LIST OF FIGURES FIG. 1 Flow Diagram of Cemented Carbide Manufacture

METHOD OF PRODUCING URANIUM

DOEpatents

Foster, L.S.; Magel, T.T.

1958-05-13

A modified process is described for the production of uranium metal by means of a bomb reduction of UF/sub 4/. Difficulty is sometimes experienced in obtaining complete separation of the uranium from the slag when the process is carried out on a snnall scale, i.e., for the production of 10 grams of U or less. Complete separation may be obtained by incorporating in the reaction mixture a quantity of MnCl/sub 2/, so that this compound is reduced along with the UF/sub 4/ . As a result a U--Mn alloy is formed which has a melting point lower than that of pure U, and consequently the metal remains molten for a longer period allowing more complete separation from the slag.

SHEATHING URANIUM

DOEpatents

Colbeck, E.W.

1959-02-01

A method is deseribed for forming a conveniently handled corrosion resistant U articlc comprising pouring molten U into an open-ended corrosion resistant metal eontainer such as Cu and its alloys, Al, or austenitic Ni stainless steel. The exposed surface of the cast U is covered with a metallic packing material such as a brazing flux consisting of Al-Si alloy. The container is sealed iii contact with substantially the entire exposed surface of the packing material. The article is then worked mechanically to reduce the cross section. l3651 A thorium--carbon alloy containing 0.1 to 0.5% by weight carbon, whieh is more resistant to water corrosion than pure thorium metal is presented. The alloy is prepared by fusing thorium metal with the desired amount of carbon at a temperature of about 1850 C. It is found that the carbon is present in the alloy as thorium monocarbide

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

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

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding andmore » Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.« less

Extrusion-formed uranium-2. 4 wt % article with decreased linear thermal expansion and method for making the same. [Patent application

DOEpatents

Anderson, R.C.; Jones, J.M.; Kollie, T.G.

1982-05-24

The present invention is directed to the fabrication of an article of uranium-2.4 wt % niobium alloy in which the linear thermal expansion in the direction transverse to the extrusion direction is less than about 0.98% between 22 and 600/sup 0/C which corresponds to a value greater than the 1.04% provided by previous extrusion operations over the same temperature range. The article with the improved thermal expansion possesses a yield strength at 0.2% offset of at least 400 MPa, an ultimate tensile strength of 1050 MPa, a compressive yield strength of at least 0.2% offset of at least 675 MPa, and an elongation of at least 25% over 25.4 mm/s. To provide this article with the improved thermal expansion, the uranium alloy billet is heated to 630/sup 0/C and extruded in the alpha phase through a die with a reduction ratio of at least 8.4:1 at a ram speed no greater than 6.8 mm/s. These critical extrusion parameters provide the article with a desired decrease in the linear thermal expansion while maintaining the selected mechanical properties without encountering crystal disruption in the article.

Extrusion-formed uranium-2.4 wt. % article with decreased linear thermal expansion and method for making the same

DOEpatents

Anderson, Robert C.; Jones, Jack M.; Kollie, Thomas G.

1982-01-01

The present invention is directed to the fabrication of an article of uranium-2.4 wt. % niobium alloy in which the linear thermal expansion in the direction transverse to the extrusion direction is less than about 0.98% between 22.degree. C. and 600.degree. C. which corresponds to a value greater than the 1.04% provided by previous extrusion operations over the same temperature range. The article with the improved thermal expansion possesses a yield strength at 0.2% offset of at least 400 MPa, an ultimate tensile strength of 1050 MPa, a compressive yield strength of at least 0.2% offset of at least 675 MPa, and an elongation of at least 25% over 25.4 mm/sec. To provide this article with the improved thermal expansion, the uranium alloy billet is heated to 630.degree. C. and extruded in the alpha phase through a die with a reduction ratio of at least 8.4:1 at a ram speed no greater than 6.8 mm/sec. These critical extrusion parameters provide the article with the desired decrease in the linear thermal expansion while maintaining the selected mechanical properties without encountering crystal disruption in the article.

Coating method for graphite

DOEpatents

Banker, John G.; Holcombe, Jr., Cressie E.

1977-01-01

A method of limiting carbon contamination from graphite ware used in induction melting of uranium alloys is provided comprising coating the graphite surface with a suspension of Y.sub.2 O.sub.3 particles in water containing about 1.5 to 4% by weight sodium carboxymethylcellulose.

Coating method for graphite

DOEpatents

Banker, J.G.; Holcombe, C.E. Jr.

1975-11-06

A method of limiting carbon contamination from graphite ware used in induction melting of uranium alloys is provided. The graphite surface is coated with a suspension of Y/sub 2/O/sub 3/ particles in water containing about 1.5 to 4 percent by weight sodium carboxymethylcellulose.

Infrared spectra and quantum chemical calculations of the uranium-carbon molecules UC, CUC, UCH, and U(CC)2

NASA Astrophysics Data System (ADS)

Wang, Xuefeng; Andrews, Lester; Ma, Dongxia; Gagliardi, Laura; Gonçalves, António P.; Pereira, Cláudia C. L.; Marçalo, Joaquim; Godart, Claude; Villeroy, Benjamin

2011-06-01

Laser evaporation of carbon rich uranium/carbon alloy targets into condensing argon or neon matrix samples gives weak infrared absorptions that increase on annealing, which can be assigned to new uranium carbon bearing species. New bands at 827.6 cm-1 in solid argon or 871.7 cm-1 in neon become doublets with mixed carbon 12 and 13 isotopes and exhibit the 1.0381 carbon isotopic frequency ratio for the UC diatomic molecule. Another new band at 891.4 cm-1 in argon gives a three-band mixed isotopic spectrum with the 1.0366 carbon isotopic frequency ratio, which is characteristic of the anti-symmetric stretching vibration of a linear CUC molecule. No evidence was found for the lower energy cyclic U(CC) isomer. Other bands at 798.6 and 544.0 cm-1 are identified as UCH, which has a uranium-carbon triple bond similar to that in UC. Evidence is found for bicyclic U(CC)2 and tricyclic U(CC)3. This work shows that U and C atoms react spontaneously to form the uranium carbide U≡C and C≡U≡C molecules with uranium-carbon triple bonds.

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

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

Diamond, D. J.; Baek, J. S.; Hanson, A. L.

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in anmore » aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.« less

Liquid uranium alloy-helium fission reactor

DOEpatents

Minkov, V.

1984-06-13

This invention describes a nuclear fission reactor which has a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200 to 1800/sup 0/C range, and even higher to 2500/sup 0/C.

Determination of boron in uranium aluminum silicon alloy by spectrophotometry and estimation of expanded uncertainty in measurement

NASA Astrophysics Data System (ADS)

Ramanjaneyulu, P. S.; Sayi, Y. S.; Ramakumar, K. L.

2008-08-01

Quantification of boron in diverse materials of relevance in nuclear technology is essential in view of its high thermal neutron absorption cross section. A simple and sensitive method has been developed for the determination of boron in uranium-aluminum-silicon alloy, based on leaching of boron with 6 M HCl and H 2O 2, its selective separation by solvent extraction with 2-ethyl hexane 1,3-diol and quantification by spectrophotometry using curcumin. The method has been evaluated by standard addition method and validated by inductively coupled plasma-atomic emission spectroscopy. Relative standard deviation and absolute detection limit of the method are 3.0% (at 1 σ level) and 12 ng, respectively. All possible sources of uncertainties in the methodology have been individually assessed, following the International Organization for Standardization guidelines. The combined uncertainty is calculated employing uncertainty propagation formulae. The expanded uncertainty in the measurement at 95% confidence level (coverage factor 2) is 8.840%.

Method of increasing the phase stability and the compressive yield strength of uranium-1 to 3 wt. % zirconium alloy

DOEpatents

Anderson, Robert C.

1986-01-01

A uranium-1 to 3 wt. % zirconium alloy characterized by high strength, high ductility and stable microstructure is fabricated by an improved thermal mechanical process. A homogenous ingot of the alloy which has been reduced in thickness of at least 50% in the two-step forging operation, rolled into a plate with a 75% reduction and then heated in vacuum at a temperature of about 750.degree. to 850.degree. C. and then quenched in water is subjected to further thermal-mechanical operation steps to increase the compressive yield strength approximately 30%, stabilize the microstructure, and decrease the variations in mechanical properties throughout the plate is provided. These thermal-mechanical steps are achieved by cold rolling the quenched plate to reduce the thickness thereof about 8 to 12%, aging the cold rolled plate at a first temperature of about 325.degree. to 375.degree. C. for five to six hours and then aging the plate at a higher temperature ranging from 480.degree. to 500.degree. C. for five to six hours prior to cooling the billet to ambient conditions and sizing the billet or plate into articles provides the desired increase in mechanical properties and phase stability throughout the plate.

Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

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

Okuniewski, Maria A.; Ganapathy, Varsha; Hamilton, Brenden

2016-09-01

ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated tomore » 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.« less

THE GRAVIMETRIC DETERMINATION OF MOLYBDENUM IN URANIUM-MOLYBDENUM ALLOYS

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

None

1959-03-01

The sample is dissolved in nitric and hydrochloric acids. After heating the solution with sulfuric acid, molybodenum is precipitated as the benzoin-oxime complex which is ignited to molybdic oxide. This is dissolved in ammonia, and the molybdenum is precipitated and weighed as lead molybdate. (auth)

A modified Embedded-Atom Method interatomic potential for uranium-silicide

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

Beeler, Benjamin; Baskes, Michael; Andersson, David

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel bene ts from higher thermal conductivity and higher ssile density compared to uranium dioxide (UO 2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling e orts are underway to address this gap in knowledge. In this study, a semi-empirical modi ed Embedded-Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential is ttedmore » to the formation energy, defect energies and structural properties of U 3Si 2. The primary phase of interest (U 3Si 2) is accurately described over a wide temperature range and displays good behavior under irradiation and with free surfaces. The potential can also describe a variety of U-Si phases across the composition spectrum.« less

A modified Embedded-Atom Method interatomic potential for uranium-silicide

DOE PAGES

Beeler, Benjamin; Baskes, Michael; Andersson, David; ...

2017-08-18

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel bene ts from higher thermal conductivity and higher ssile density compared to uranium dioxide (UO 2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling e orts are underway to address this gap in knowledge. In this study, a semi-empirical modi ed Embedded-Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential is ttedmore » to the formation energy, defect energies and structural properties of U 3Si 2. The primary phase of interest (U 3Si 2) is accurately described over a wide temperature range and displays good behavior under irradiation and with free surfaces. The potential can also describe a variety of U-Si phases across the composition spectrum.« less

A modified Embedded-Atom Method interatomic potential for uranium-silicide

NASA Astrophysics Data System (ADS)

Beeler, Benjamin; Baskes, Michael; Andersson, David; Cooper, Michael W. D.; Zhang, Yongfeng

2017-11-01

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel benefits from higher thermal conductivity and higher fissile density compared to uranium dioxide (UO2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling efforts are underway to address this gap in knowledge. In this study, a semi-empirical modified Embedded-Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential is fitted to the formation energy, defect energies and structural properties of U3Si2. The primary phase of interest (U3Si2) is accurately described over a wide temperature range and displays good behavior under irradiation and with free surfaces. The potential can also describe a variety of U-Si phases across the composition spectrum.

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

Stillman, J. A.; Feldman, E. E.; Wilson, E. H.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains themore » results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo (U-10Mo).« 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

High-temperature electrically conductive ceramic composite and method for making same

DOEpatents

Beck, David E.; Gooch, Jack G.; Holcombe, Jr., Cressie E.; Masters, David R.

1983-01-01

The present invention relates to a metal-oxide ceramic composition useful in induction heating applications for treating uranium and uranium alloys. The ceramic composition is electrically conductive at room temperature and is nonreactive with molten uranium. The composition is prepared from a particulate admixture of 20 to 50 vol. % niobium and zirconium oxide which may be stabilized with an addition of a further oxide such as magnesium oxide, calcium oxide, or yttria. The composition is prepared by blending the powders, pressing or casting the blend into the desired product configuration, and then sintering the casting or compact in an inert atmosphere. In the casting operation, calcium aluminate is preferably added to the admixture in place of a like quantity of zirconia for providing a cement to help maintain the integrity of the sintered product.

Volume collapse phase transitions in cerium-praseodymium alloys under high pressure

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

Perreault, Christopher S.; Velisavljevic, Nenad; Samudrala, Gopi K.

Cerium-12at%Praseodymium(Ce 0.88Pr 0.12) and Ce-50at%Praseodymium(Ce 0.50Pr 0.50) alloy samples that contain a random solid-solution of Ce (4f1 (J=5/2)) and Pr (4f2 (J=4)) localized f-states have been studied by angle dispersive x-ray diffraction in a diamond anvil cell to a pressure of 65 GPa and 150 GPa respectively using a synchrotron source. Ce 0.88Pr 0.12 alloy crystallizes in a face-centered cubic (γ-phase) structure at ambient conditions, while Ce 0.50Pr 0.50 alloy crystallizes in the double hexagonal close packed (dhcp) structure at ambient conditions. Two distinct volume collapse transitions are observed in Ce 0.88Pr 0.12 alloy at 1.5 GPa and 18 GPamore » with volume change of 8.5% and 3% respectively. In contrast, Ce 0.50Pr 0.50 alloy shows only a single volume collapse of 5.6% at 20 GPa on phase transformation to α-Uranium structure under high pressure. Electrical transport measurements under high pressure show anomalies in electrical resistance at phase transitions for both compositions of this alloy.« less

Volume collapse phase transitions in cerium-praseodymium alloys under high pressure

DOE PAGES

Perreault, Christopher S.; Velisavljevic, Nenad; Samudrala, Gopi K.; ...

2018-06-08

Cerium-12at%Praseodymium(Ce 0.88Pr 0.12) and Ce-50at%Praseodymium(Ce 0.50Pr 0.50) alloy samples that contain a random solid-solution of Ce (4f1 (J=5/2)) and Pr (4f2 (J=4)) localized f-states have been studied by angle dispersive x-ray diffraction in a diamond anvil cell to a pressure of 65 GPa and 150 GPa respectively using a synchrotron source. Ce 0.88Pr 0.12 alloy crystallizes in a face-centered cubic (γ-phase) structure at ambient conditions, while Ce 0.50Pr 0.50 alloy crystallizes in the double hexagonal close packed (dhcp) structure at ambient conditions. Two distinct volume collapse transitions are observed in Ce 0.88Pr 0.12 alloy at 1.5 GPa and 18 GPamore » with volume change of 8.5% and 3% respectively. In contrast, Ce 0.50Pr 0.50 alloy shows only a single volume collapse of 5.6% at 20 GPa on phase transformation to α-Uranium structure under high pressure. Electrical transport measurements under high pressure show anomalies in electrical resistance at phase transitions for both compositions of this alloy.« less

HEU Holdup Measurements on 321-M A-Lathe

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

Dewberry, R.A.

The Analytical Development Section of SRTC was requested by the Facilities Disposition Division (FDD) of the Savannah River Site to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The results of the holdup assays are essential for determining compliance with the solid waste Waste Acceptance Criteria, Material Control and Accountability, and to meet criticality safety controls. Three measurement systems were used to determine highly enrichedmore » uranium (HEU) holdup. This report covers holdup measurements on the A-Lathe that was used to machine uranium-aluminum-alloy (U-Al). Our results indicated that the lathe contained more than the limits stated in the Waste Acceptance Criteria (WAC) for the solid waste E-Area Vaults. Thus the lathe was decontaminated three times and assayed four times in order to bring the amounts of uranium to an acceptable content. This report will discuss the methodology, Non-Destructive Assay (NDA) measurements, and results of the U-235 holdup on the lathe.« less

11. AERIAL VIEW LOOKING NORTH AT THE BUILDING 800 ...

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

11. AERIAL VIEW LOOKING NORTH AT THE BUILDING 800 - AREA COMPLEX. ENRICHED URANIUM COMPONENTS WERE MANUFACTURED IN THIS AREA OF THE SITE. BUILDING 881, IN THE RIGHT FOREGROUND OF THE PHOTOGRAPH, WAS THE ORIGINAL PLANT B. BUILDING 883, USED FOR ROLLING AND FORMING URANIUM COMPONENTS, IS DIRECTLY TO THE NORTH OF BUILDING 881. TO THE EAST OF BUILDING 883 IS BUILDING 885, A RESEARCH AND DEVELOPMENT FACILITY FOR ALLOYS AND NON-PLUTONIUM METALS. IN THE FOREGROUND TO THE WEST OF BUILDING 881 IS AN OFFICE BUILDING, 850 (6/7/90). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2013 CFR

2013-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2012 CFR

2012-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2011 CFR

2011-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

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

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth

Grain boundaries in metallic alloys often play a crucial role, not only in determining the mechanical properties or thermal stability of alloys, but also in dictating the phase transformation kinetics during thermomechanical processing. We demonstrate that locally stabilized structure and compositional segregation at grain boundaries—“grain boundary complexions”—in a complex multicomponent alloy can be modified to influence the kinetics of cellular transformation during subsequent thermomechanical processing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography analysis of a metallic nuclear fuel highly relevant to worldwide nuclear non-proliferation efforts —uranium-10 wt% molybdenum (U-10Mo) alloy, new evidence for the existence of grainmore » boundary complexion is provided. We then modified the concentration of impurities dissolved in Υ-UMo grain interiors and/or segregated to Υ-UMo grain boundaries by changing the homogenization treatment, and these effects were used used to retard the kinetics of cellular transformation during subsequent sub-eutectoid annealing in this U-10-Mo alloy during sub-eutectoid annealing. Thus, this work provided insights on tailoring the final microstructure of the U-10Mo alloy, which can potentially improve the irradiation performance of this important class of alloy fuels.« less

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Irradiation of three T-111 clad uranium nitride fuel pins for 8070 hours at 990 C (1815 F)

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.; Gedeon, L.; Galbo, R. J.

1973-01-01

The design and successful operation of three tantalum alloy (Ta-8W-2Hf) clad uranium mononitride (UN) fuel pins irradiated for 8070 hr at 990 C (1815 F) is described. Two pin diameters having measured burnups of 0.47 and 0.90 uranium atom percent were tested. No clad failures or swelling was detected; however, postirradiation clad samples tested failed with 1 percent strain. The fuel density decrease was 2 percent, and the fission gas release was less than 0.05 percent. Isotropic fuel swelling, which averaged about 0.5 percent, was less than fuel pin assembly clearances. Thus the clad was not strained. Thermocouples with a modified hot zone operated at average temperatures to 1100 C (2012 F) without failure. Factors that influence the ability to maintain uniform clad temperature as well as the results of the heat transfer calculations are discussed.

METALLURGY DIVISION QUARTERLY REPORT FOR JULY, AUGUST, AND SEPTEMBER 1957

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

None

1958-10-01

Advanced Water Reactor Program. Three firings were made of initial closed-porosity fuel pellet bodies. Each firing coatained pellets of the composition 90 wt.% ThO/sub 2/-10 wt.%fl U0/sub 2/ with various additives and firing variables. Fast Power Breeder Reactor Program. To determine the potential usefulness of a Zr-5 wt. % Pu alloy, the fabricability of the alloy was tested. The manufacture of rod stock from which fuel and blanket elements for the Mark III loading of the EBR-1 were prcduced has been completed. The effect of irradiation on extruded and heat-treated U-2 wt.% Zr alloy for the EBR- 1 is reported.more » Fabrication procedures for making graphite-U/sub 3/O/sub 8/ test specimens for the TREAT Reactor were investigated. Advanced Engineering and Development. Ultrasonic bond tests were conducted on 590 EBR-1 Mark III blanket fuel elemeats. The blanket rods and part of the fuel rcds for the EBR-1 Mark III loading are being checked for cladding thickness by an eddy current system. Investigations of corrosionresistant Zr-Nb alloy were coatinued. Corrosion of MR alloys is being studied Ln support of the Mighty Mouse reactor program. Dynamic corrosion tests were performed on aluminum alloys, and results are included. Prcduction, Treatment, and Properties of Materials. The progress of the program of preparing highpurity Pu by fused salt electrolysis is summarized. Velocities of ultrasonic waves propagated in directions suitable for determining the room- temperature elastic moduli C/sub 12/, C/sub 13/, and C/sub 23/ of alpha U were determined. investigation of recrystallization in heavily coldrolled alpha- uranium sheet without a texture change was essentially concluded during this quarter. Selfdiffasion runs in polycrystalline uranium in the gamma phase, using the sputtering technique, have yielded a tentative value for the diffusion coefficient between 10/sup -8/ and 10/sup -7/ cm/sup 2/second. The preparation of high-purity U-Pan alloys is reponted. The data for the alpha-tobeta transformation temperatures in high-purity U and U-C alloys were confirmed by experiments on new specimens. Microstructure, density, and thermal arrest data were obtained for several injection cast, nominal U-5 wt.%fl fissium and U-8 wt.%fl fissium alloys. Phase diagrams are preseated for U-Mo and U-Ru alloys. Alloy Theory and The Nature of Solids. Four new isomorphs of Ti/sub 2/Ni have been discovered. Effects of Irradiation on Materials. The experimental and analytical work on the radial distribution of thermal neutrons within cylindrically shaped fuel specimens during irradiation was completed. (For preceding period see ANL-5790.) (W.L.H.)« less

Defect structures induced by high-energy displacement cascades in γ uranium

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

Miao, Yinbin; Beeler, Benjamin; Deo, Chaitanya

Displacement cascade simulations were conducted for the c uranium system based on molecular dynamics. A recently developed modified embedded atom method (MEAM) potential was employed to replicate the atomic interactions while an embedded atom method (EAM) potential was adopted to help characterize the defect structures induced by the displacement cascades. The atomic displacement process was studied by providing primary knock-on atoms (PKAs) with kinetic energies from 1 keV to 50 keV. The influence of the PKA incident direction was examined. The defect structures were analyzed after the systems were fully relaxed. The states of the self-interstitial atoms (SIAs) were categorizedmore » into various types of dumbbells, the crowdion, and the octahedral interstitial. The voids were determined to have a polyhedral shape with {110} facets. The size distribution of the voids was also obtained. The results of this study not only expand the knowledge of the microstructural evolution in irradiated c uranium, but also provide valuable references for the radiation-induced defects in uranium alloy fuels.« less

State Environmental Policy Act (SEPA) environmental checklist forms for 304 Concretion Facility Closure Plan. Revision 2

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

Not Available

The 300 Area of the Hanford Site contains reactor fuel manufacturing facilities and several research and development laboratories. 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 304 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 304 Facility is now undergoing closure asmore » defined in the Resource Conservation and Recovery Act (RCRA) of 1976 and the Washington Administrative Code (WAC) Dangerous Waste Regulations, WAC 173-303-040. This closure plan presents a description of the 304 Facility, the history of materials and waste managed, and the procedures that will be followed to close the 304 Facility. The 304 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

Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2

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

Coffey, Greg W.; Meinhardt, Kerry D.; Joshi, Vineet V.

2015-03-01

The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce amore » uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is that plating can be performed under conditions that would greatly reduce the quantity of intermetallics that form at the interface between the zirconium and U-10Mo; unlike roll bonding, the molten salt plating approach would allow for complete coverage of the U-10Mo foil with zirconium. When utilizing the experimental parameters developed for zirconium plating onto molybdenum, a uranium fluoride reaction product was formed at the Zr/U-10Mo interface. By controlling the initial plating potential, the uranium fluoride could be prevented; however, the targeted zirconium thickness (25 ±12.5 μm) could not be achieved while maintaining 100% coverage.« less

Oxygen potential of uranium--plutonium oxide as determined by controlled- atmosphere thermogravimetry

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

Swanson, Gerald C.

1975-10-01

The oxygen-to-metal atom ratio, or O/M, of solid solution uranium- plutonium oxide reactor fuel is a measure of the concentration of crystal defects in the oxide which affect many fuel properties, particularly, fuel oxygen potential. Fabrication of a high-temperature oxygen electrode, employing an electro-active tip of oxygen-deficient solid-state electrolyte, intended to confirm gaseous oxygen potentials is described. Uranium oxide and plutonium oxide O/M reference materials were prepared by in situ oxidation of high purity metals in the thermobalance. A solid solution uranium-plutonium oxide O/M reference material was prepared by alloying the uranium and plutonium metals in a yttrium oxide cruciblemore » at 1200°C and oxidizing with moist He at 250°C. The individual and solid solution oxides were isothermally equilibrated with controlled oxygen potentials between 800 and 1300°C and the equilibrated O/ M ratios calculated with corrections for impurities and buoyancy effects. Use of a reference oxygen potential of -100 kcal/mol to produce an O/M of 2.000 is confirmed by these results. However, because of the lengthy equilibration times required for all oxides, use of the O/M reference materials rather than a reference oxygen potential is recommended for O/M analysis methods calibrations.« less

Interim Report on Mixing During the Casting of LEU-10Mo Plates in the Triple Plate Molds

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

Aikin, Jr., Robert M.

LEU-10%Mo castings are commonly produced by down blending unalloyed HEU with a DU-12.7%Mo master-alloy. This work uses process modeling to provide insight into the mixing of the unalloyed uranium and U-Mo master alloy during melting and mold filling of a triple plate casting. Two different sets of situations are considered: (1) mixing during mold filling from a compositionally stratified crucible and (2) convective mixing of a compositionally stratified crucible during mold heating. The mold filling simulations are performed on the original Y-12 triple plate mold and the horizontal triple plate mold.

Inclusions and Substructures in Uranium of Nuclear Purity. Report No. 51; INCLUSIONES Y SUBESTRUCTURAS EN URANIO DE PUREZA NUCLEAR. INFORME NO. 51

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

Biloni, H.; Lindenvald, N.; Sabato, J.A.

1961-01-01

The inclusions in uranium of nuclear purity (UC, UH/sub 3/, UO/sub 2/, UO, UN, and the complexes which include the intersolubility of U with C and N or with C, N, and O) were . analyzed metallographically, and the results reported by other authors were discussed critically. The existence of the fine precipitate reticular substructure, sensitive to thermal treatments, which generally appears in uraniunn was analyzed. Its origins were discussed in accordance with bibliographic data. Complementary data for its comprehension are given from the metallographic analysis of U--Al and U-- Fe alloys with low Al and Fe concentrations. (tr-auth)

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

Thermal Conductivity of Metallic Uranium

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

Hin, Celine

This project has developed a modeling and simulation approaches to predict the thermal conductivity of metallic fuels and their alloys. We focus on two methods. The first method has been developed by the team at the University of Wisconsin Madison. They developed a practical and general modeling approach for thermal conductivity of metals and metal alloys that integrates ab-initio and semi-empirical physics-based models to maximize the strengths of both techniques. The second method has been developed by the team at Virginia Tech. This approach consists of a determining the thermal conductivity using only ab-initio methods without any fitting parameters. Bothmore » methods were complementary. The models incorporated both phonon and electron contributions. Good agreement with experimental data over a wide temperature range were found. The models also provided insight into the different physical factors that govern the thermal conductivity under different temperatures. The models were general enough to incorporate more complex effects like additional alloying species, defects, transmutation products and noble gas bubbles to predict the behavior of complex metallic alloys like U-alloy fuel systems under burnup. 3 Introduction Thermal conductivity is an important thermal physical property affecting the performance and efficiency of metallic fuels [1]. Some experimental measurement of thermal conductivity and its correlation with composition and temperature from empirical fitting are available for U, Zr and their alloys with Pu and other minor actinides. However, as reviewed in by Kim, Cho and Sohn [2], due to the difficulty in doing experiments on actinide materials, thermal conductivities of metallic fuels have only been measured at limited alloy compositions and temperatures, some of them even being negative and unphysical. Furthermore, the correlations developed so far are empirical in nature and may not be accurate when used for prediction at conditions far from those used in the original fitting. Moreover, as fuels burn up in the reactor and fission products are built up, thermal conductivity is also significantly changed [3]. Unfortunately, fundamental understanding of the effect of fission products is also currently lacking. In this project, we probe thermal conductivity of metallic fuels with ab initio calculations, a theoretical tool with the potential to yield better accuracy and predictive power than empirical fitting. This work will both complement experimental data by determining thermal conductivity in wider composition and temperature ranges than is available experimentally, and also develop mechanistic understanding to guide better design of metallic fuels in the future. So far, we focused on α-U perfect crystal, the ground-state phase of U metal. We focus on two methods. The first method has been developed by the team at the University of Wisconsin Madison. They developed a practical and general modeling approach for thermal conductivity of metals and metal alloys that integrates ab-initio and semi-empirical physics-based models to maximize the strengths of both techniques. The second method has been developed by the team at Virginia Tech. This approach consists of a determining the thermal conductivity using only ab-initio methods without any fitting parameters. Both methods were complementary and very helpful to understand the physics behind the thermal conductivity in metallic uranium and other materials with similar characteristics. In Section I, the combined model developed at UWM is explained. In Section II, the ab-initio method developed at VT is described along with the uranium pseudo-potential and its validation. Section III is devoted to the work done by Jianguo Yu at INL. Finally, we will present the performance of the project in terms of milestones, publications, and presentations.« less

Interdiffusion behaviors in doped molybdenum uranium and aluminum or aluminum silicon dispersion fuels: Effects of the microstructure

NASA Astrophysics Data System (ADS)

Allenou, J.; Tougait, O.; Pasturel, M.; Iltis, X.; Charollais, F.; Anselmet, M. C.; Lemoine, P.

2011-09-01

Si addition to Al is considered as a promising route to reduce (U,Mo)-Al interaction kinetics, due to its accumulation in the interaction layer, yielding the formation of silicide phases. The (U,Mo) alloy microstructure, and especially its homogenization state, could play a role on this accumulation process. The addition of a third element in γ(U,Mo) could also influence diffusion mechanisms of Al and Si. These two parameters were studied by means of diffusion couple experiments by joining γU based alloys with Al and (Al,Si) alloy. Chemical elements X added into γ(U,Mo) were thoroughly chosen on the following criteria: (i) the potential solubility of the alloying element into the γ(U,Mo) matrix, (ii) its capability to form the ternary aluminides based on the CeCr 2Al 20 and Ho 6Mo 4Al 43 - types, and (iii) the feasibility to control the microstructure of the alloys. On this basis, a test matrix is defined. It concerns γ(U80,Mo15,X5) alloys (in at.%) with X = Y, Cu, Zr, Ti or Cr. These alloys were homogenized and coupled with Al or (Al,Si) alloy. Results evidenced, first, the importance of the state of homogenization of the γ(U,Mo) binary alloy on interaction processes with (Al,Si) alloy, and the benefit on the diffusion of Si through the interaction layer, as observed on the elementary concentration profiles, when the third element X has some solubility into γ(U,Mo) alloy.

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy

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

Prabhakaran, Ramprashad; Joshi, Vineet V.; Rhodes, Mark A.

2016-10-01

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy

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

Prabhakaran, Ramprashad; Joshi, Vineet V.; Rhodes, Mark A.

2016-03-30

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

Phase separation of metal-added corium and its effect on a steam explosion

NASA Astrophysics Data System (ADS)

Min, B. T.; Kim, J. H.; Hong, S. W.; Hong, S. H.; Park, I. K.; Song, J. H.; Kim, H. D.

2008-07-01

To simulate a relocation of molten core material and its interaction phenomenon with water during a severe accident in a nuclear reactor, a typical corium of UO 2/ZrO 2/Zr/Stainless steel mixed at a 62 wt%, 15 wt%, 12 wt% and 11 wt%, respectively, was melted and then cooled down to become a solidified ingot. It was shown that the molten corium was separated into two layers, of which the upper layer was oxide mixtures and the lower layer was metal alloys. The upper layer was UO 2 and ZrO 2 and the lower layer mostly consisted of metal mixtures such as uranium, zirconium and stainless steel. Iron content varied with the positions and about a half of it existed as an alloy such as Fe 2U. Uranium metal was produced by reduction of UO 2 by zirconium metal. The average densities of the upper oxide layer and the lower metal layer were 8.802 and 9.411 g/cm 3, respectively. In another test, metal-added molten corium was poured into water and it showed that a steam explosion could occur by applying an external trigger.

Performance of a Steel/Oxide Composite Waste Form for Combined Waste Steams from Advanced Electrochemical Processes

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

Indacochea, J. E.; Gattu, V. K.; Chen, X.

The results of electrochemical corrosion tests and modeling activities performed collaboratively by researchers at the University of Illinois at Chicago and Argonne National Laboratory as part of workpackage NU-13-IL-UIC-0203-02 are summarized herein. The overall objective of the project was to develop and demonstrate testing and modeling approaches that could be used to evaluate the use of composite alloy/ceramic materials as high-level durable waste forms. Several prototypical composite waste form materials were made from stainless steels representing fuel cladding, reagent metals representing metallic fuel waste streams, and reagent oxides representing oxide fuel waste streams to study the microstructures and corrosion behaviorsmore » of the oxide and alloy phases. Microelectrodes fabricated from small specimens of the composite materials were used in a series of electrochemical tests to assess the corrosion behaviors of the constituent phases and phase boundaries in an aggressive acid brine solution at various imposed surface potentials. The microstructures were characterized in detail before and after the electrochemical tests to relate the electrochemical responses to changes in both the electrode surface and the solution composition. The results of microscopic, electrochemical, and solution analyses were used to develop equivalent circuit and physical models representing the measured corrosion behaviors of the different materials pertinent to long-term corrosion behavior. This report provides details regarding (1) the production of the composite materials, (2) the protocol for the electrochemical measurements and interpretations of the responses of multi-phase alloy and oxide composites, (3) relating corrosion behaviors to microstructures of multi-phase alloys based on 316L stainless steel and HT9 (410 stainless steel was used as a substitute) with added Mo, Ni, and/or Mn, and (4) modeling the corrosion behaviors and rates of several alloy/oxide composite materials made with added lanthanide and uranium oxides. These analyses show the corrosion behaviors of the alloy/ceramic composite materials are very similar to the corrosion behaviors of multi-phase alloy waste forms, and that the presence of oxide inclusions does not impact the corrosion behaviors of the alloy phases. Mixing with metallic waste streams is beneficial to lanthanide and uranium oxides in that they react with Zr in the fuel waste to form highly durable zirconates. The measured corrosion behaviors suggest properly formulated composite materials would be suitable waste forms for combined metallic and oxide waste streams generated during electrometallurgical reprocessing of spent nuclear fuel. Electrochemical methods are suitable for evaluating the durability and modeling long-term behavior of composite waste forms: the degradation model developed for metallic waste forms can be applied to the alloy phases formed in the composite and an affinity-based mineral dissolution model can be applied to the ceramic phases.« less

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

NASA Technical Reports Server (NTRS)

Bowles, K. J.; Gluyas, R. E.

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

WELDING PROCESS

DOEpatents

Zambrow, J.; Hausner, H.

1957-09-24

A method of joining metal parts for the preparation of relatively long, thin fuel element cores of uranium or alloys thereof for nuclear reactors is described. The process includes the steps of cleaning the surfaces to be jointed, placing the sunfaces together, and providing between and in contact with them, a layer of a compound in finely divided form that is decomposable to metal by heat. The fuel element members are then heated at the contact zone and maintained under pressure during the heating to decompose the compound to metal and sinter the members and reduced metal together producing a weld. The preferred class of decomposable compounds are the metal hydrides such as uranium hydride, which release hydrogen thus providing a reducing atmosphere in the vicinity of the welding operation.

Examination of T-111 clad uranium nitride fuel pins irradiated up to 13,000 hours at a clad temperature of 990 C

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.

1973-01-01

The examination of 27 fuel pins irradiated for up to 13,000 hours at 990 C is described. The fuel pin clad was a tantalum alloy with uranium nitride as the nuclear fuel. Two nominal fuel pin diameters were tested with a maximum burnup of 2.34 atom percent. Twenty-two fuel pins were tested for fission gas leaks; thirteen pins leaked. Clad ductility tests indicated clad embrittlement. The embrittlement is attributed to hydrogen from an n,p reaction in the fuel. Fuel swelling was burnup dependent, and the amount of fission gas release was low, generally less than 0.5 percent. No incompatibilities between fuel, liner, and clad were in evidence.

Update on specified European R and D efforts

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

Not Available

1980-10-01

Information was collected for DOE on various European research programs of interest: Shell-Koppers coal gasification demonstration plant, fluidized-bed combustion pilot plant, a boiler super heat system, energy conservation on ships, waste heat utilization from large diesel engines and nuclear power plants and uranium enrichment plants, coal-water slurries with additive (CARBOGEL), electrostatic precipitators, radial inflow turbines, carbonization, heat pumps, heat exchangers, gas turbines, and research on heat resisting alloys and corrosion protection of these alloys. A number of organizations expressed a desire for creation of a formal interchange with DOE on specific subjects of mutual interest (one organization is unhappy aboutmore » furnishing information to DOE). (LTN)« less

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

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

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

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

10 CFR 110.22 - General license for the export of source material.

Code of Federal Regulations, 2010 CFR

2010-01-01

... country not listed in § 110.28: (1) Uranium or thorium, other than U-230, U-232, Th-227, and Th-228, in any substance in concentrations of less than 0.05 percent by weight. (2) Thorium, other than Th-227 and Th-228, in incandescent gas mantles or in alloys in concentrations of 5 percent or less. (3) Th...

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

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

Montgomery, Robert O.; Bennett, Wendy D.; Henager, Charles H.

2016-09-12

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impactmore » of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.« less

The Influence of Oxygen and Sulfur on Uranium Partitioning Into the Core

NASA Astrophysics Data System (ADS)

Moore, R. D., Jr.; Van Orman, J. A.; Hauck, S. A., II

2017-12-01

Uranium, along with K and Th, may provide substantial long-term heating in planetary cores, depending on the magnitude of their partitioning into the metal during differentiation. In general, non-metallic light elements are known to have a large influence on the partitioning of trace elements, and the presence of sulfur is known to enhance the partitioning of uranium into the metal. Data from the steelmaking literature indicate that oxygen also enhances the solubility of oxygen in liquid iron alloys. Here we present experimental data on the partitioning of U between immiscible liquids in the Fe-S-O system, and use these data along with published metal-silicate partitioning data to calibrate a quantitative activity model for U in the metal. We also determined partition coefficients for Th, K, Nb, Nd, Sm, and Yb, but were unable to fully constrain activity models for these elements with available data. A Monte Carlo fitting routine was used to calculate U-S, U-O, and U-S-O interaction coefficients, and their associated uncertainties. We find that the combined interaction of uranium with sulfur and oxygen is predominant, with S and O together enhancing the solubility of uranium to a far greater degree than either element in isolation. This suggests that uranium complexes with sulfite or sulfate species in the metal. For a model Mars core composition containing 14 at% S and 5 at% O, the metal/silicate partition coefficient for U is predicted to be an order of magnitude larger than for a pure Fe-Ni core.

Environmental acceptability of high-performance alternatives for depleted uranium penetrators

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

Kerley, C.R.; Easterly, C.E.; Eckerman, K.F.

1996-08-01

The Army`s environmental strategy for investigating material substitution and management is to measure system environmental gains/losses in all phases of the material management life cycle from cradle to grave. This study is the first in a series of new investigations, applying material life cycle concepts, to evaluate whether there are environmental benefits from increasing the use of tungsten as an alternative to depleted uranium (DU) in Kinetic Energy Penetrators (KEPs). Current military armor penetrators use DU and tungsten as base materials. Although DU alloys have provided the highest performance of any high-density alloy deployed against enemy heavy armor, its low-levelmore » radioactivity poses a number of environmental risks. These risks include exposures to the military and civilian population from inhalation, ingestion, and injection of particles. Depleted uranium is well known to be chemically toxic (kidney toxicity), and workplace exposure levels are based on its renal toxicity. Waste materials containing DU fragments are classified as low-level radioactive waste and are regulated by the Nuclear Regulatory Commission. These characteristics of DU do not preclude its use in KEPs. However, long-term management challenges associated with KEP deployment and improved public perceptions about environmental risks from military activities might be well served by a serious effort to identify, develop, and substitute alternative materials that meet performance objectives and involve fewer environmental risks. Tungsten, a leading candidate base material for KEPS, is potentially such a material because it is not radioactive. Tungsten is less well studied, however, with respect to health impacts and other environmental risks. The present study is designed to contribute to the understanding of the environmental behavior of tungsten by synthesizing available information that is relevant to its potential use as a penetrator.« less

Finite-element model to predict roll-separation force and defects during rolling of U-10Mo alloys

NASA Astrophysics Data System (ADS)

Soulami, Ayoub; Burkes, Douglas E.; Joshi, Vineet V.; Lavender, Curt A.; Paxton, Dean

2017-10-01

A major goal of the Convert Program of the U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA) is to enable high-performance research reactors to operate with low-enriched uranium rather than the high-enriched uranium currently used. To this end, uranium alloyed with 10 wt% molybdenum (U-10Mo) represents an ideal candidate because of its stable gamma phase, low neutron caption cross section, acceptable swelling response, and predictable irradiation behavior. However, because of the complexities of the fuel design and the need for rolled monolithic U-10Mo foils, new developments in processing and fabrication are necessary. This study used a finite-element code, LS-DYNA, as a predictive tool to optimize the rolling process. Simulations of the hot rolling of U-10Mo coupons encapsulated in low-carbon steel were conducted following two different schedules. Model predictions of the roll-separation force and roll pack thicknesses at different stages of the rolling process were compared with experimental measurements. The study reported here discussed various attributes of the rolled coupons revealed by the model (e.g., waviness and thickness non-uniformity like dog-boning). To investigate the influence of the cladding material on these rolling defects, other cases were simulated: hot rolling with alternative can materials, namely, 304 stainless steel and Zircaloy-2, and bare-rolling. Simulation results demonstrated that reducing the mismatch in strength between the coupon and can material improves the quality of the rolled sheet. Bare-rolling simulation results showed a defect-free rolled coupon. The finite-element model developed and presented in this study can be used to conduct parametric studies of several process parameters (e.g., rolling speed, roll diameter, can material, and reduction).

Compact power reactor

DOEpatents

Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

Carcinogenicity and Immunotoxicity of Embedded Depleted Uranium and Heavy-Metal Tungsten Alloy in Rodents

DTIC Science & Technology

2006-10-01

the implantation site. In addition, WA-implanted rats (high-dose group) exhibited splenomegaly and hematological changes suggesting polycythemia ...controls. The hematological changes observed in the high-dose WA rats are suggestive of polycythemia . Cobalt has been used experimentally to induce... polycythemia in rats (Rakusan et al., 2001; Endoh et al., 2000) although the concentration required is far greater than found in the WA pellets. The

Liquid uranium alloy-helium fission reactor

DOEpatents

Minkov, Vladimir

1986-01-01

This invention teaches a nuclear fission reactor having a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200.degree.-1800.degree. C. range, and even higher to 2500.degree. C., limited only by the thermal effectiveness of the structural materials, increasing the efficiency of power generation from the normal 30-35% with 300.degree.-500.degree. C. upper limit temperature to 50-65%. Irradiation of the circulating liquid fuel, as contrasted to only localized irradiation of a solid fuel, provides improved fuel utilization.

The influence of impurities on the crystal structure and mechanical properties of additive manufactured U–14at.% Nb

DOE PAGES

Wu, Amanda S.; Brown, Donald W.; Clausen, Bjørn; ...

2017-03-01

Uranium-niobium alloys can exist with significantly different microstructures and mechanical properties, heavily influenced by thermomechanical processing history and impurities. In this study, the influence of Ti and other impurities is studied on uranium-14 at.% niobium additively manufactured using laser powder bed fusion. In two different metallic impurity levels were investigated and a Nb equivalent (Nbeq) composition is defined to represent the impurities. Furthermore, in-situ neutron diffraction during compression loading shows that increased Nbeq promotes the formation of γ°-tetragonal phase at the expense of α''-monoclinic phase, resulting in 2 × higher yield strength than water quenched α'' and a strain inducedmore » transformation to α'' with superelastic strains to 4.5%.« less

Evaluation of tantalum-alloy-clad uranium mononitride fuel specimens from 7500-hour, 1040 C pumped-lithium-loop test

NASA Technical Reports Server (NTRS)

Watson, G. K.

1974-01-01

Simulated nuclear fuel element specimens, consisting of uranium mononitride (UN) fuel cylinders clad with tungsten-lined T-111, were exposed for up to 7500 hr at 1040 C (1900 F) in a pumped-lithium loop. The lithium flow velocity was 1.5 m/sec (5 ft/sec) in the specimen test section. No evidence of any compatibility problems between the specimens and the flowing lithium was found based on appearance, weight change, chemistry, and metallography. Direct exposure of the UN to the lithium through a simulated cladding crack resulted in some erosion of the UN in the area of the defect. The T-111 cladding was ductile after lithium exposure, but it was sensitive to hydrogen embrittlement during post-test handling.

The influence of impurities on the crystal structure and mechanical properties of additive manufactured U–14at.% Nb

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

Wu, Amanda S.; Brown, Donald W.; Clausen, Bjørn

Uranium-niobium alloys can exist with significantly different microstructures and mechanical properties, heavily influenced by thermomechanical processing history and impurities. In this study, the influence of Ti and other impurities is studied on uranium-14 at.% niobium additively manufactured using laser powder bed fusion. In two different metallic impurity levels were investigated and a Nb equivalent (Nbeq) composition is defined to represent the impurities. Furthermore, in-situ neutron diffraction during compression loading shows that increased Nbeq promotes the formation of γ°-tetragonal phase at the expense of α''-monoclinic phase, resulting in 2 × higher yield strength than water quenched α'' and a strain inducedmore » transformation to α'' with superelastic strains to 4.5%.« less

Mechanical Behavior of Additively Manufactured Uranium-6 wt. pct. Niobium

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

Wu, A. S.; Wraith, M. W.; Burke, S. C.

This report describes an effort to process uranium-6 weight% niobium using laser powder bed fusion. The chemistry, crystallography, microstructure and mechanical response resulting from this process are discussed with particular emphasis on the effect of the laser powder bed fusion process on impurities. In an effort to achieve homogenization and uniform mechanical behavior from different builds, as well as to induce a more conventional loading response, we explore post-processing heat treatments on this complex alloy. Elevated temperature heat treatment for recrystallization is evaluated and the effect of recrystallization on mechanical behavior in laser powder bed fusion processed U-6Nb is discussed.more » Wrought-like mechanical behavior and grain sizes are achieved through post-processing and are reported herein.« less

Effect of thermo-mechanical processing on microstructure and mechanical properties of U - Nb - Zr alloys: Part 2 - U - 3 wt % Nb - 9 wt % Zr and U - 9 wt% Nb - 3 wt% Zr

NASA Astrophysics Data System (ADS)

Morais, Nathanael Wagner Sales; Lopes, Denise Adorno; Schön, Cláudio Geraldo

2018-04-01

The present work is the second and final part of an extended investigation on Usbnd Nb - Zr alloys. It investigates the effect of mechanical processing routes on microstructure of alloys U - 3 wt % Nb - 9 wt % Zr and U - 9 wt% Nb - 3 wt% Zr, through X-ray diffraction and scanning electron microscopy, completing the investigation, which started with alloy U - 6 wt% Nb - 6 wt% Zr in part 1. Mechanical properties are determined using microhardness and bending tests and correlated with the developed microstructures. The results show that processing sequence, in particular the inclusion of a 1000 °C heat treatment step, affects significantly the microstructure and mechanical properties of these alloys alloy in different ways. Microstructural characterization shows that both alloys present significant volume fraction of precipitates of a body-centered cubic (BCC) γ-Nb-Zr rich phase in addition the uranium-rich matrix. Bending tests show that sample ductility does not correlate necessarily with hardness and that the key factor appears to be the amount of the γ-Nb-Zr precipitates, which controls the matrix microstructure. Samples with a monoclinic α″ cellular microstructure and/or with the tetragonally-distorted BCC phase (γ0), although not strictly ductile, showed the largest allowed strains-before-break and complete elastic recovery of the broken pieces, pointing out to the macroscopic observation of superelasticity.

Interfacial Chemical Interactions in the (Alumina/Graphite/Al Alloys) System: Thermodynamic Modeling and Experimental Results

NASA Astrophysics Data System (ADS)

Gelbstein, M.; Edry, I.; Froumin, N.; Frage, N.

2009-04-01

The stability of alumina-coated graphite couples in liquid Al is investigated in the 1373 to 1573 K temperature range. A thermodynamic model was carried out to determine the mechanisms controlling the couple stability and the effect of alloying Al with high melting point element for instance U (up to 3 at. pct). It was established that the dissolved uranium dose not play any role in the interfacial interactions and that the couple stability is governed by the interactions with Al resulting in the release of gaseous products. The experiments focused on wetting kinetics under conditions allowing for an in-situ reduction of the alumina coating by the liquid Al. The experimental results confirm the predictions of the thermodynamic analysis.

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

Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

DOE PAGES

Marshall, Margaret A.

2014-11-04

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with themore » GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.« less

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

Processing and Applications of Depleted Uranium Alloy Products

DTIC Science & Technology

1976-09-01

temperal,,r at the wheel-metal interface, thus tending to produce surface cracks and in some cases to burn the metal. Data on speeds and feeds inr...comprehensive current resource of technical information on the development and utilization of advcnlod metal- or ceramic-base materials. The Center is operated...under the sponsorship of the Department of Defense. Neither the United Staxes Government nor any person acting on be ilf of the United States Government

Corrosion Assessment of Candidate Materials for the SHINE Subcritical Assembly Vessel and Components FY15 Report

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

Pawel, Steven J.

2016-01-01

In the previous report of this series, a literature review was performed to assess the potential for substantial corrosion issues associated with the proposed SHINE process conditions to produce 99Mo. Following the initial review, substantial laboratory corrosion testing was performed emphasizing immersion and vapor-phase exposure of candidate alloys in a wide variety of solution chemistries and temperatures representative of potential exposure conditions. Stress corrosion cracking was not identified in any of the exposures up to 10 days at 80°C and 10 additional days at 93°C. Mechanical properties and specimen fracture face features resulting from slow-strain rate tests further supported amore » lack of sensitivity of these alloys to stress corrosion cracking. Fluid velocity was found not to be an important variable (0 to ~3 m/s) in the corrosion of candidate alloys at room temperature and 50°C. Uranium in solution was not found to adversely influence potential erosion-corrosion. Potentially intense radiolysis conditions slightly accelerated the general corrosion of candidate alloys, but no materials were observed to exhibit an annualized rate above 10 μm/y.« less

Genotoxic Changes to Rodent Cells Exposed in Vitro to Tungsten, Nickel, Cobalt and Iron

PubMed Central

Bardack, Stephanie; Dalgard, Clifton L.; Kalinich, John F.; Kasper, Christine E.

2014-01-01

Tungsten-based materials have been proposed as replacements for depleted uranium in armor-penetrating munitions and for lead in small-arms ammunition. A recent report demonstrated that a military-grade composition of tungsten, nickel, and cobalt induced a highly-aggressive, metastatic rhabdomyosarcoma when implanted into the leg muscle of laboratory rats to simulate a shrapnel wound. The early genetic changes occurring in response to embedded metal fragments are not known. In this study, we utilized two cultured rodent myoblast cell lines, exposed to soluble tungsten alloys and the individual metals comprising the alloys, to study the genotoxic effects. By profiling cell transcriptomes using microarray, we found slight, yet distinct and unique, gene expression changes in rat myoblast cells after 24 h metal exposure, and several genes were identified that correlate with impending adverse consequences of ongoing exposure to weapons-grade tungsten alloy. These changes were not as apparent in the mouse myoblast cell line. This indicates a potential species difference in the cellular response to tungsten alloy, a hypothesis supported by current findings with in vivo model systems. Studies examining genotoxic-associated gene expression changes in cells from longer exposure times are warranted. PMID:24619124

Microstructure Characterization of RERTR Fuel

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

J. Gan; B. D. Miller; D. D. Keiser

2008-09-01

A variety of phases have the potential to develop in the irradiated fuels for the reduced enrichment research test reactor (RERTR) program. To study the radiation stability of these potential phases, three depleted uranium alloys were cast. The phases of interest were identified including U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, UAl4, and U6Mo4Al43. These alloys were irradiated with 2.6 MeV protons at 200ºC up to 3.0 dpa. The microstructure is characterized using SEM and TEM. Microstructural characterization for an archive dispersion fuel plate (U-7Mo fuel particles in Al-2%Si cladding) was also carried out. TEM sample preparation for the irradiated dispersion fuel has beenmore » developed.« less

Isoelectronic substitutions and aluminium alloying in the Ta-Nb-Hf-Zr-Ti high-entropy alloy superconductor

NASA Astrophysics Data System (ADS)

von Rohr, Fabian O.; Cava, Robert J.

2018-03-01

High-entropy alloys (HEAs) are a new class of materials constructed from multiple principal elements statistically arranged on simple crystallographic lattices. Due to the large amount of disorder present, they are excellent model systems for investigating the properties of materials intermediate between crystalline and amorphous states. Here we report the effects of systematic isoelectronic replacements, using Mo-Y, Mo-Sc, and Cr-Sc mixtures, for the valence electron count 4 and 5 elements in the body-centered cubic (BCC) Ta-Nb-Zr-Hf-Ti high-entropy alloy (HEA) superconductor. We find that the superconducting transition temperature Tc strongly depends on the elemental makeup of the alloy, and not exclusively its electron count. The replacement of niobium or tantalum by an isoelectronic mixture lowers the transition temperature by more than 60%, while the isoelectronic replacement of hafnium, zirconium, or titanium has a limited impact on Tc. We further explore the alloying of aluminium into the nearly optimal electron count [TaNb] 0.67(ZrHfTi) 0.33 HEA superconductor. The electron count dependence of the superconducting Tc for (HEA)Al x is found to be more crystallinelike than for the [TaNb] 1 -x(ZrHfTi) x HEA solid solution. For an aluminum content of x =0.4 the high-entropy stabilization of the simple BCC lattice breaks down. This material crystallizes in the tetragonal β -uranium structure type and superconductivity is not observed above 1.8 K.

Bedrock Prime: How Can the United States Best Address the Need to Achieve Dominance within the Subterranean Domain?

DTIC Science & Technology

2012-12-14

monolithic construction (formed from a single piece of hard material) and made of a high-strength alloy. The use of monolithic construction eliminates...is the class of sub-caliber solid depleted uranium or tungsten “darts” that are conveyed by a sabot during gun launching. Current penetrating...and deeper fortifications which are largely immune to the current United States weapons inventory. Advanced construction and design techniques

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

DOEpatents

Swanson, J.L.

1961-07-11

The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

Carcinogenicity and Immunotoxicity of Embedded Depleted Uranium and Heavy-Metal Tungsten Alloy in Rodents

DTIC Science & Technology

2006-10-01

local metal- induced toxicity daily for two weeks following surgery and weekly thereafter for the duration of the study. Task 1 - Determine whether...A quick and simple method for the quantitation of lactate dehydrogenase release in measurements of cellular cytotoxicity and tumor necrosis factor...4 months, no signs of local or systemic toxic- ity were observed (Peuster et al. 2003). Studies on health effects of Ni and Co are more numerous

High Velocity Firings of Slug Projectiles in a Double-Travel 120-MM Gun System

DTIC Science & Technology

1991-04-01

constraints presented by TBD. This charge configuration was then tested using aluminium slug projectiles to avoid the unnecessary expenditure of APFSDS...test projectile was a depleted uranium alloy (U-.75Ti) rod with a standard, four piece, aluminum sabot assembly. The launch package had a nominal...the rod is shown in Figure 2. Figure 2. Scaled, Long Rod Penetrator. Figure 3. Aluminium Slug Projectile. The aluminium slug rounds, fired at Range 18

Manufacturing Experience for Oxide Dispersion Strengthened Alloys

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

Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.

2016-09-22

This report documents the results of the development and the manufacturing experience gained at the Pacific Northwest National Laboratories (PNNL) while working with the oxide dispersion strengthened (ODS) materials MA 956, 14YWT, and 9YWT. The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. ODS materials have the potential to provide improved performance for the U-Mo concept.

Interdiffusion behavior of U3Si2 with FeCrAl via diffusion couple studies

NASA Astrophysics Data System (ADS)

Hoggan, Rita E.; He, Lingfeng; Harp, Jason M.

2018-04-01

Uranium silicide (U3Si2) is a candidate to replace uranium oxide (UO2) as light water reactor (LWR) fuel because of its higher thermal conductivity and higher fissile density relative to the current standard, UO2. A class of Fe, Cr, Al alloys collectively known as FeCrAl alloys that have superior mechanical and oxidation resistance are being considered as an alternative to the standard Zirconium based LWR cladding. The interdiffusion behavior between FeCrAl and U3Si2 is investigated in this study. Commercially available FeCrAl, along with U3Si2 pellets were placed in diffusion couples. Individual tests were ran at temperatures ranging from 500 °C to 1000 °C for 30 h and 100 h. The interdiffusion was analyzed with an optical microscope, scanning electron microscope, and transmission electron microscope. Uniform and planar interdiffusion layers along the material interface were illustrated with backscatter electron micrographs and energy-dispersive X-ray spectroscopy. Electron diffraction was used to validate phases present in the system, including distinct U2Fe3Si/UFe2 and UFeSi layers at the material interface. U and Fe diffused far into the FeCrAl and U3Si2 matrix, respectively, in the higher temperature tests. No interaction was observed at 500 °C for 30 h.

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

DOE PAGES

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth; ...

2018-03-30

Here, we demonstrate here that locally stabilized structure and compositional segregation at grain boundaries in a complex multicomponent alloy can be modified using high temperature homogenization treatment to influence the kinetics of phase transformations initiating from grain boundaries during subsequent low temperature annealing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography of a model multicomponent metallic alloy —uranium-10 wt% molybdenum (U-10Mo) a nuclear fuel, that is highly relevant to worldwide nuclear non-proliferation efforts, we demonstrate the ability to change the structure and compositional segregation at grain boundary, which then controls the subsequent discontinuous precipitation kinetics during sub-eutectoid annealing.more » A change in grain boundary from one characterized by segregation of Mo and impurities at grain boundary to a phase boundary with a distinct U 2MoSi 2C wetting phase precipitates introducing Ni and Al rich interphase complexions caused a pronounced reduction in area fraction of subsequent discontinuous precipitation. The broader implication of this work is in highlighting the role of grain boundary structure and composition in metallic alloys on dictating the fate of grain boundary initiated phase transformations like discontinuous precipitation or cellular transformation. This work highlights a new pathway to tune the grain boundary structure and composition to tailor the final microstructure of multicomponent metallic alloys.« less

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

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

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth

Here, we demonstrate here that locally stabilized structure and compositional segregation at grain boundaries in a complex multicomponent alloy can be modified using high temperature homogenization treatment to influence the kinetics of phase transformations initiating from grain boundaries during subsequent low temperature annealing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography of a model multicomponent metallic alloy —uranium-10 wt% molybdenum (U-10Mo) a nuclear fuel, that is highly relevant to worldwide nuclear non-proliferation efforts, we demonstrate the ability to change the structure and compositional segregation at grain boundary, which then controls the subsequent discontinuous precipitation kinetics during sub-eutectoid annealing.more » A change in grain boundary from one characterized by segregation of Mo and impurities at grain boundary to a phase boundary with a distinct U 2MoSi 2C wetting phase precipitates introducing Ni and Al rich interphase complexions caused a pronounced reduction in area fraction of subsequent discontinuous precipitation. The broader implication of this work is in highlighting the role of grain boundary structure and composition in metallic alloys on dictating the fate of grain boundary initiated phase transformations like discontinuous precipitation or cellular transformation. This work highlights a new pathway to tune the grain boundary structure and composition to tailor the final microstructure of multicomponent metallic alloys.« less

Evidence of mixed valence states in U M2Al 3 ( M = Ni, Pd) studied by X-ray photoemission spectroscopy

NASA Astrophysics Data System (ADS)

Fujimori, Shin-ichi; Saito, Yasuharu; Sato, Noriaki; Komatsubara, Takemi; Suzuki, Shoji; Sato, Shigeru; Ishii, Takehiko

1998-01-01

We have measured the XPS valence band and core-level spectra of U M2Al 3 ( M = Ni and Pd). The results are compared with those of reference materials, dilute alloy U 0.1La 0.9Pd 2Al 3 and itinerant 5 f compound URh 3. The similarity of the core-level spectra between UPd 2Al 3 and U 0.1La 0.9Pd 2Al 3 suggests that their core-level spectra are governed by the interaction between U 5 f and ligand states of neighboring palladium and aluminum sites, with negligible contributions from neighboring uranium states. A complex satellite structure, observed in the core-level spectra of U M2Al 3, suggests that the uranium atoms are in the strong mixed valence states with 5 f2(U 4+) and 5 f3(U 3+).

REMOVAL OF CHLORIDE FROM AQUEOUS SOLUTIONS

DOEpatents

Hyman, M.L.; Savolainen, J.E.

1960-01-01

A method is given for dissolving reactor fuel elements in which the uranium is associated with a relatively inert chromium-containing alloy such as stainless steel. An aqueous mixture of acids comprising 2 to 2.5 molar hydrochloric acid and 4 to 8 molar nitric acid is employed in dissolving the fuel element. In order io reduce corrosion in subsequent processing of the resulting solution, chloride values are removed from the solution by contacting it with concentrated nitric acid at an elevated temperature.

Process to separate transuranic elements from nuclear waste

DOEpatents

Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

1989-03-21

A process is described for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

Diffusion Couple Alloying of Refractory Metals in Austenitic and Ferritic/Martensitic Steels

DTIC Science & Technology

2012-03-01

applications of austenitic stainless steel and ferritic/martensitic steel can vary from structural and support components in the reactor core to reactor fuel ... fuel . It serves as a boundary to prevent both fission products from escaping to the core coolant, and segregates the fuel from the coolant to...uranium oxide (UO2) fuel in the core . It resists corrosion by the fuel matrix on the inner surface of the cladding and the liquid sodium coolant on

Unraveling the Age Hardening Response in U-Nb Alloys

DOE PAGES

Hackenberg, Robert Errol; Hemphill, Geralyn M. Sewald; Forsyth, Robert Thomas; ...

2016-11-15

Complicating factors that have stymied understanding of uranium-niobium’s aging response are briefly reviewed, including (1) niobium inhomogeneity, (2) machining damage effects on tensile properties, (3) early-time transients of ductility increase, and (4) the variety of phase transformations. A simple Logistic-Arrhenius model was applied to predict yield and ultimate tensile strengths and tensile elongation of U-4Nb as a function of thermal age. Lastly, fits to each model yielded an apparent activation energy that was compared with phase transformation mechanisms.

Process to separate transuranic elements from nuclear waste

DOEpatents

Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

1988-07-12

A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

PROGRESS ON THE STUDY OF THE URANIUM-ALUMINUM-IRON CONSTITUTION DIAGRAM FOR THE PERIOD APRIL 1-AUGUST 31, 1963

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

Russell, R.B.

Progress is reported of a research program on deterinining the U--Al-- Fe constitution diagram up to approximates 1000 ppm each of Al and Fe. Results are reported of metallographic examinations of U--Al, U--Fe, and U--Al--Fe alloys, and tentative phase diagrams are given for all three systems. Lattice parameters and electrical resistances were measured in an effort to determine the solid solubilities of Al and Fe in U. (D.L.C.)

Process to separate transuranic elements from nuclear waste

DOEpatents

Johnson, Terry R.; Ackerman, John P.; Tomczuk, Zygmunt; Fischer, Donald F.

1989-01-01

A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

Surface characterization of U(AlxSi1-x)3 alloy and its interaction with O2 and H2O, at room temperature

NASA Astrophysics Data System (ADS)

Matmor, M.; Cohen, S.; Rafailov, G.; Vaknin, M.; Shamir, N.; Gouder, T.; Zalkind, S.

2018-02-01

Surface characterization and the interactions of U(AlxSi1-x)3 alloy (x = 0.57) with oxygen and water vapor were studied, utilizing X-Ray Photoelectron Spectroscopy and Direct Recoil Spectrometry, at room temperature. The U 4f spectrum of U(AlxSi1-x)3 alloy exhibits weak correlation satellites, suggesting an itinerant description of the U 5f states for this compound. The Al and Si 2p lines are chemically shifted to lower binding energies. Exposing the alloy to oxygen and water vapor results in oxidation of mainly the uranium and aluminum components, while silicon is only slightly oxidized. Oxygen was found to be a stronger oxidizer than water vapor and the trend is consistent with the more negative enthalpies of formation of metal oxides produced by the O2 reaction, as compared to H2O. During oxygen exposure, fast oxidation occurs by oxide islands nucleation and lateral growth, followed by oxidation of the sub-surface, up to ∼4 nm, at 1000 L exposure. Water initially reacts with the surface by full dissociation and oxide islands formation, which is then covered by hydroxides. Only a minor increase in the oxide thickness of up to ∼2.5 nm, was observed after coalescence.

First-principles study of the surface properties of U-Mo system

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

Mei, Zhi-Gang; Liang, Linyun; Yacout, Abdellatif M.

U-Mo alloys are promising fuels for future high-performance research reactors with low enriched uranium. Surface properties, such as surface energy, are important inputs for mesoscale simulations (e.g., phase field method) of fission gas bubble behaviors in irradiated nuclear fuels. The lack of surface energies of U-Mo alloys prevents an accurate modeling of the morphology of gas bubbles and gas bubble-induced fuel swelling. To this end, we study the surface properties of U-Mo system, including bcc Mo, alpha-U, gamma-U, and gamma U-Mo alloys. All surfaces up to a maximum Miller index of three and two are calculated for cubic Mo andmore » gamma-U and non-cubic alpha-U, respectively. The equilibrium crystal shapes of bcc Mo, alpha-U and gamma-U are constructed using the calculated surface energies. The dominant surface orientations and the area fraction of each facet are determined from the constructed equilibrium crystal shape. The disordered gamma U-Mo alloys are simulated using the Special Quasirandom Structure method. The (1 1 0) and (1 0 0) surface energies of gamma U-7Mo and U-10Mo alloys are predicted to lie between those of gamma-U and bcc Mo, following a linear combination of the two constituents' surface energies. To better compare with future measurements of surface energies, the area fraction weighted surface energies of alpha-U, gamma-U and gamma U-7Mo and U-10Mo alloys are also predicted. (C) 2017 Published by Elsevier B.V.« less

Irradiation performance of U-Mo monolithic fuel

DOE PAGES

Meyer, M. K.; Gan, J.; Jue, J. F.; ...

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

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

M.K. Meyer; J. Gan; J.-F. Jue

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

Preliminary neutronics design of china lead-alloy cooled demonstration reactor (CLEAR-III) for nuclear waste transmutation

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

Chen, Z.; Southwest Science and Technology Univ., No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031; Chen, Y.

2012-07-01

China Lead-Alloy cooled Demonstration Reactor (CLEAR-III), which is the concept of lead-bismuth cooled accelerator driven sub-critical reactor for nuclear waste transmutation, was proposed and designed by FDS team in China. In this study, preliminary neutronics design studies have primarily focused on three important performance parameters including Transmutation Support Ratio (TSR), effective multiplication factor and blanket thermal power. The constraint parameters, such as power peaking factor and initial TRU loading, were also considered. In the specific design, uranium-free metallic dispersion fuel of (TRU-Zr)-Zr was used as one of the CLEAR-III fuel types and the ratio between MA and Pu was adjustedmore » to maximize transmutation ratio. In addition, three different fuel zones differing in the TRU fraction of the fuel were respectively employed for this subcritical reactor, and the zone sizes and TRU fractions were determined such that the linear powers of these zones were close to each other. The neutronics calculations and analyses were performed by using Multi-Functional 4D Neutronics Simulation System named VisualBUS and nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library). In the preliminary design, the maximum TSRLLMA was {approx}11 and the blanket thermal power was {approx}1000 MW when the effective multiplication factor was 0.98. The results showed that good performance of transmutation could be achieved based on the subcritical reactor loaded with uranium-free fuel. (authors)« less

U-Mo Monolithic Fuel for Nuclear Research and Test Reactors

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

Prabhakaran, Ramprashad

The metallic fuel selected to replace the current HEU fuels in the research and test reactors is the LEU-10 weight % Mo alloy in the form of a thin sheet or foil encapsulated in AA6061 aluminum alloy with a zirconium interlayer. In order to effectively lead this pursuit, new developments in processing and fabrication of the fuel elements have been initiated, along with a better understanding of material behavior before and after irradiation as a result of these new developments. This editorial note gives an introduction about research and test reactors, need for HEU to LEU conversion, fuel requirements, highmore » uranium density monolithic fuel development and an overview of the four articles published in the December 2017 issue of JOM under a special topic titled “U-Mo Monolithic Fuel for Nuclear Research and Test Reactors”.« less

Rolling Process Modeling Report. Finite-Element Model Validation and Parametric Study on various Rolling Process parameters

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

Soulami, Ayoub; Lavender, Curt A.; Paxton, Dean M.

2015-06-15

Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum alloy plate-type fuel for high-performance research reactors in the United States. This work supports the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization Reactor Conversion Program. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll-separating forces for various rolling mill geometries for PNNL, Babcock & Wilcox Co., Y-12 National Security Complex, Los Alamos National Laboratory, and Idaho National Laboratory. The model developed and presented in a previous report has been subjected to further validationmore » study using new sets of experimental data generated from a rolling mill at PNNL. Simulation results of both hot rolling and cold rolling of uranium-10% molybdenum coupons have been compared with experimental results. The model was used to predict roll-separating forces at different temperatures and reductions for five rolling mills within the National Nuclear Security Administration Fuel Fabrication Capability project. This report also presents initial results of a finite-element model microstructure-based approach to study the surface roughness at the interface between zirconium and uranium-10% molybdenum.« less

Thermodynamics of La and U and the separation factor of U/La in fused Me(Ga-40 wt.% In)/3LiCl-2KCl system

NASA Astrophysics Data System (ADS)

Smolenski, Valeri; Novoselova, Alena; Volkovich, Vladimir A.

2017-11-01

Separation of lanthanides and actinides can be achieved in a unique "molten chloride - liquid metal" system. Electrode potentials were recorded vs. Cl-/Cl2 reference electrode and the temperature dependencies of the apparent standard potentials of La-(Ga-In) and U-(Ga-In) alloys were determined. Thermodynamic properties and separation factor of lanthanum and uranium were calculated. The obtained data show the perspective for using this system in future innovation method for recovery of nuclear waste.

Canned pump having a high inertia flywheel

DOEpatents

Veronesi, Luciano; Raimondi, ALbert A.

1989-01-01

A canned pump is described which includes a motor, impeller, shaft, and high inertia flywheel mounted within a hermetically sealed casing. The flywheel comprises a heavy metal disk made preferably of a uranium alloy with a stainless steel shell sealably enclosing the heavy metal. The outside surfaces of the stainless steel comprise thrust runners and a journal for mating with, respectively, thrust bearing shoes and radial bearing segments. The bearings prevent vibration of the pump and, simultaneously, minimize power losses normally associated with the flywheel resulting from frictionally pumping surrounding fluid.

Canned pump having a high inertia flywheel

DOEpatents

Veronesi, L.; Raimondi, A.A.

1989-12-12

A canned pump is described which includes a motor, impeller, shaft, and high inertia flywheel mounted within a hermetically sealed casing. The flywheel comprises a heavy metal disk made preferably of a uranium alloy with a stainless steel shell sealably enclosing the heavy metal. The outside surfaces of the stainless steel comprise thrust runners and a journal for mating with, respectively, thrust bearing shoes and radial bearing segments. The bearings prevent vibration of the pump and, simultaneously, minimize power losses normally associated with the flywheel resulting from frictionally pumping surrounding fluid. 5 figs.

Superlattice Multinanolayered Thin Films of SiO2/SiO2 + Ge for Thermoelectric Device Applications

DTIC Science & Technology

2013-04-05

radioiso- tope sources in the past. In a space nuclear reactor system, the energy source is the heat generated by the controlled fission of uranium ...to the nanodots and/or nanocluster formations in the multilayered thin films. This is one of the expected results of the ion beam bombardments on...very large (150 W m 1 K 1 for Si and 63 W m 1 K 1 for Ge). The lattice thermal conductivity can be substantially reduced by alloy formation between

The interaction between nitride uranium and stainless steel

NASA Astrophysics Data System (ADS)

Shornikov, D. P.; Nikitin, S. N.; Tarasov, B. A.; Baranov, V. G.; Yurlova, M. S.

2016-04-01

Uranium nitride is most popular nuclear fuel for Fast Breeder Reactor New Generation. In-pile experiments at reactor BOR-60 was shown an interaction between nitride fuel and stainless steel in the range of 8-11% burn up (HA). In order to investigate this interaction has been done diffusion tests of 200 h and has been shown that the reaction occurs in the temperature range 1000-1100 ° C. UN interacted with steel in case of high pollution oxygen (1000-2000 ppm). Also has been shown to increase interaction UN with EP-823 steel in the presence of cesium. In this case the interaction layer had a thickness about 2-3 μm. Has been shown minimal interaction with new ODS steel EP-450. The interaction layer had a thickness less then 2 μm. Did not reveal the influence of tellurium and iodine increased interaction. It was show compatibility at 1000 °C between UN and EP-450 ODS steel, chrome steel, alloying aluminium and silicium.

Submersion criticality safety of tungsten-rhenium urania cermet fuel for space propulsion and power applications

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

A.E. Craft; R. C. O'Brien; S. D. Howe

Nuclear thermal rockets are the preferred propulsion technology for a manned mission to Mars, and tungsten–uranium oxide cermet fuels could provide significant performance and cost advantages for nuclear thermal rockets. A nuclear reactor intended for use in space must remain subcritical before and during launch, and must remain subcritical in launch abort scenarios where the reactor falls back to Earth and becomes submerged in terrestrial materials (including seawater, wet sand, or dry sand). Submersion increases reflection of neutrons and also thermalizes the neutron spectrum, which typically increases the reactivity of the core. This effect is typically very significant for compact,more » fast-spectrum reactors. This paper provides a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor with a range of fuel compositions. Each submersion case considers both the rhenium content in the matrix alloy and the uranium oxide volume fraction in the cermet. The inclusion of rhenium significantly improves the submersion criticality safety of the reactor. While increased uranium oxide content increases the reactivity of the core, it does not significantly affect the submersion behavior of the reactor. There is no significant difference in submersion behavior between reactors with rhenium distributed within the cermet matrix and reactors with a rhenium clad in the coolant channels. The combination of the flooding of the coolant channels in submersion scenarios and the presence of a significant amount of spectral shift absorbers (i.e. high rhenium concentration) further decreases reactivity for short reactor cores compared to longer cores.« less

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

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

Montierth, Leland M.

2016-07-19

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element designmore » for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.« less

Role of Si on the Diffusional Interactions Between U-Mo and Al-Si Alloys at 823 K (550 °C)

NASA Astrophysics Data System (ADS)

Perez, Emmanuel; Sohn, Yong-Ho; Keiser, Dennis D.

2013-01-01

U-Mo dispersions in Al-alloy matrix and monolithic fuels encased in Al-alloy are under development to fulfill the requirements for research and test reactors to use low-enriched molybdenum stabilized uranium alloy fuels. Significant interaction takes place between the U-Mo fuel and Al during manufacturing and in-reactor irradiation. The interaction products are Al-rich phases with physical and thermal characteristics that adversely affect fuel performance and result in premature failure. Detailed analysis of the interdiffusion and microstructural development of this system was carried through diffusion couples consisting of U-7 wt pct Mo, U-10 wt pct Mo and U-12 wt pct Mo in contact with pure Al, Al-2 wt pct Si, and Al-5 wt pct Si, annealed at 823 K (550 °C) for 1, 5 and 20 hours. Scanning electron microscopy and transmission electron microscopy were employed for the analysis. Diffusion couples consisting of U-Mo in contact with pure Al contained UAl3, UAl4, U6Mo4Al43, and UMo2Al20 phases. Additions of Si to the Al significantly reduced the thickness of the interdiffusion zone. The interdiffusion zones developed Al- and Si-enriched regions, whose locations and size depended on the Si and Mo concentrations in the terminal alloys. In these couples, the (U,Mo)(Al,Si)3 phase was observed throughout the interdiffusion zone, and the U6Mo4Al43 and UMo2Al20 phases were observed only where the Si concentrations were low.

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

An interdiffusional model for prediction of the interaction layer growth in the system uranium molybdenum/aluminum

NASA Astrophysics Data System (ADS)

Soba, A.; Denis, A.

2007-03-01

The codes PLACA and DPLACA, elaborated in this working group, simulate the behavior of a plate-type fuel containing in its core a foil of monolithic or dispersed fissile material, respectively, under normal operation conditions of a research reactor. Dispersion fuels usually consist of ceramic particles of a uranium compound in a high thermal conductivity matrix. The use of particles of a U-Mo alloy in a matrix of Al requires especially devoted subroutines able to simulate the growth of the interaction layer that develops between the particles and the matrix. A model is presented in this work that gives account of these particular phenomena. It is based on the assumption that diffusion of U and Al through the layer is the rate-determining step. Two moving interfaces separate the growing reaction layer from the original phases. The kinetics of these boundaries are solved as Stefan problems. In order to test the model and the associated code, some previous, simpler problems corresponding to similar systems for which analytical solutions or experimental data are known were simulated. Experiments performed with planar U-Mo/Al diffusion couples are reported in the literature, which purpose is to obtain information on the system parameters. These experiments were simulated with PLACA. Results of experiments performed with U-Mo particles disperse in Al either without or with irradiation, published in the open literature were simulated with DPLACA. A satisfactory prediction of the whole reaction layer thickness and of the individual fractions corresponding to alloy and matrix consumption was obtained.

Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

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

Bretscher, M.M.; Matos, J.E.

At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U{sub 3}Si{sub 2} fuel is about 6.0 g U/cm{sup 3}. The French Commissariat a l`Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L`Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm{sup 3} and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersionmore » fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm{sup 3}. On the other hand, UN is the least reactive fuel because of the relatively large {sup 14}N(n,p) cross section. For a fixed value of k{sub eff}, the required {sup 235}U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO{sub 2} dispersions are only useful for uranium densities below 5.0 g/cm{sup 3}. In this density range, however, UO{sub 2} is more reactive than U{sub 3}Si{sub 2}.« less

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

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

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the Highmore » Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.« less

The irradiation behavior of atomized U-Mo alloy fuels at high temperature

NASA Astrophysics Data System (ADS)

Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

2001-04-01

Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

2016-05-01

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

A physical description of fission product behavior fuels for advanced power reactors.

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

Kaganas, G.; Rest, J.; Nuclear Engineering Division

2007-10-18

The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuelsmore » under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.« less

A model to predict thermal conductivity of irradiated U–Mo dispersion fuel

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

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

The Office of Materials Management and Minimization Reactor Conversion Program continues to develop existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. The program is focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layermore » formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.« less

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

Burkes, Douglas E.; Senor, David J.; Casella, Andrew M.

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO2-stainless steel dispersion fuels and used currently available thermal-mechanical property information for the materials ofmore » interest in the current proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the 235U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the yield strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of interaction layer formation and can extend the performance of a fuel plate under a certain set of irradiation conditions, primarily moderate heat flux and burnup. Increasing the dispersed fuel particle diameter may also be effective, but only when combined with other parameters, e.g., lower enrichment and increased Si concentration. The model may serve as a valuable tool in initial experimental design.« less

Development of Alternative Technetium Waste Forms

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

Czerwinski, Kenneth

2013-09-13

The UREX+1 process is under consideration for the separation of transuranic elements from spent nuclear fuel. The first steps of this process extract the fission product technicium-99 ({sup 99}Tc) into an organic phase containing tributylphosphate together with uranium. Treatment of this stream requires the separation of Tc from U and placement into a suitable waste storage form. A potential candidate waste form involves immobilizing the Tc as an alloy with either excess metallic zirconium or stainless steel. Although Tc-Zr alloys seem to be promising waste forms, alternative materials must be investigated. Innovative studies related to the synthesis and behavior ofmore » a different class of Tc materials will increase the scientific knowledge related to development of Tc waste forms. These studies will also provide a better understanding of the behavior of {sup 99}Tc in repository conditions. A literature survey has selected promising alternative waste forms for further study: technetium metallic alloys, nitrides, oxides, sulfides, and pertechnetate salts. The goals of this project are to 1) synthesize and structurally characterize relevant technetium materials that may be considered as waste forms, 2) investigate material behavior in solution under different conditions of temperature, electrochemical potential, and radiation, and 3) predict the long-term behavior of these materials.« less

Effects of the foil flatness on the stress-strain characteristics of U10Mo alloy based monolithic mini-plates

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

Hakan Ozaltun; Pavel Medvedev

The effects of the foil flatness on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation were studied. Monolithic plate-type fuels are a new fuel form being developed for research and test reactors to achieve higher uranium densities. This concept facilitates the use of low-enriched uranium fuel in the reactor. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. To evaluate the effects of the foil flatness on the stress-strain behavior of the plates during fabrication, irradiation and shutdown stages, a representative plate frommore » RERTR-12 experiments (Plate L1P756) was considered. Both fabrication and irradiation processes of the plate were simulated by using actual irradiation parameters. The simulations were repeated for various foil curvatures to observe the effects of the foil flatness on the peak stress and strain magnitudes of the fuel elements. Results of fabrication simulations revealed that the flatness of the foil does not have a considerable impact on the post fabrication stress-strain fields. Furthermore, the irradiation simulations indicated that any post-fabrication stresses in the foil would be relieved relatively fast in the reactor. While, the perfectly flat foil provided the slightly better mechanical performance, overall difference between the flat-foil case and curved-foil case was not significant. Even though the peak stresses are less affected, the foil curvature has several implications on the strain magnitudes in the cladding. It was observed that with an increasing foil curvature, there is a slight increase in the cladding strains.« less

Effects of heat treatment on U-Mo fuel foils with a zirconium diffusion barrier

NASA Astrophysics Data System (ADS)

Jue, Jan-Fong; Trowbridge, Tammy L.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.; Keiser, Dennis D.

2015-05-01

A monolith fuel design based on U-Mo alloy has been selected as the fuel type for conversion of the United States' high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U-Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U-Mo foil during fabrication alters the microstructure of both the U-10Mo fuel meat and the U-Mo/Zr interface. This work studied the effects of post-rolling annealing treatment on the microstructure of the co-rolled U-Mo fuel meat and the U-Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U-Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ∼9, ∼13, and ∼20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U-Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U-Mo coupon homogenization. The phases in the Zr/U-Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.

Effects of heat treatment on U–Mo fuel foils with a zirconium diffusion barrier

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

Jue, Jan-Fong; Trowbridge, Tammy L.; Breckenridge, Cynthia R.

A monolith fuel design based on U–Mo alloy has been selected as the fuel type for conversion of the United States’ high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U–Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U–Mo foil during fabrication alters the microstructure of both the U–10Mo fuel meat and the U–Mo/Zr interface. This work studied the effects of post-rolling annealing treatmentmore » on the microstructure of the co-rolled U–Mo fuel meat and the U–Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U–Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ~9, ~13, and ~20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U–Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U–Mo coupon homogenization. The phases in the Zr/U–Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.« less

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

Marshall, Margaret A.

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with themore » GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.« less

Characterization of deformation mechanisms in zirconium alloys: effect of temperature and irradiation

NASA Astrophysics Data System (ADS)

Long, Fei

Zirconium alloys have been widely used in the CANDU (CANada Deuterium Uranium) reactor as core structural materials. Alloy such as Zircaloy-2 has been used for calandria tubes; fuel cladding; the pressure tube is manufactured from alloy Zr-2.5Nb. During in-reactor service, these alloys are exposed to a high flux of fast neutron at elevated temperatures. It is important to understand the effect of temperature and irradiation on the deformation mechanism of zirconium alloys. Aiming to provide experimental guidance for future modeling predictions on the properties of zirconium alloys this thesis describes the result of an investigation of the change of slip and twinning modes in Zircaloy-2 and Zr-2.5Nb as a function of temperature and irradiation. The aim is to provide scientific fundamentals and experimental evidences for future industry modeling in processing technique design, and in-reactor property change prediction of zirconium components. In situ neutron diffraction mechanical tests carried out on alloy Zircaloy-2 at three temperatures: 100¢ªC, 300¢ªC, and 500¢ªC, and described in Chapter 3. The evolution of the lattice strain of individual grain families in the loading and Poisson's directions during deformation, which probes the operation of slip and twinning modes at different stress levels, are described. By using the same type of in situ neutron diffraction technique, tests on Zr-2.5Nb pressure tube material samples, in either the fast-neutron irradiated or un-irradiated condition, are reported in Chapter 4. In Chapter 5, the measurement of dislocation density by means of line profile analysis of neutron diffraction patterns, as well as TEM observations of the dislocation microstructural evolution, is described. In Chapter 6 a hot-rolled Zr-2.5Nb with a larger grain size compared with the pressure tubing was used to study the development of dislocation microstructures with increasing plastic strain. In Chapter 7, in situ loading of heavy ion irradiated hot-rolled Zr-2.5Nb alloy is described, providing evidence for the interaction between moving dislocations and irradiation induced loops. Chapter 8 gives the effect on the dislocation structure of different levels of compressive strains along two directions in the hot-rolled Zr-2.5Nb alloy. By using high resolution neutron diffraction and TEM observations, the evolution of type and dislocation densities, as well as changes of dislocation microstructure with plastic strain were characterized.

TEM analysis of irradiation-induced interaction layers in coated UMo/X/Al trilayer systems (X= Ti, Nb, Zr, and Mo)

NASA Astrophysics Data System (ADS)

Chiang, H.-Y.; Wiss, T.; Park, S.-H.; Dieste-Blanco, O.; Petry, W.

2018-02-01

Uranium-molybdenum (UMo) alloy powder embedded in an Al matrix is considered as a promising candidate for fuel conversion of research reactors. A modified system with a diffusion barrier X as coating, UMo/X/Al trilayer (X = Ti, Zr, Nb, and Mo), has been investigated to suppress interdiffusion between UMo and the Al matrix. The trilayer systems were tested by swift heavy ion irradiation, the thereby created interaction zone has been analyzed by scanning transmission electron microscopy (STEM) and energy-dispersive X-ray spectroscopy (EDX). Detailed structural characterization are presented and compared to earlier μ-XRD analysis.

Quasi-Static Compression and Tensile Stress-Strain Curves, Tantalum - 10% Tungsten and 300 Grade Maraging Steel

DTIC Science & Technology

1986-05-01

available high density metal. It was investigated as a practical substitute for ballistic tungsten or uranium alloys. It was also used as a core material... PPM ) H ɝ ɝ ɝ C ᝺ ណ ណ N 12 ɝ 7 0 41 អ 60 Si ម ម Ti ង V ង ង ង Cr ᝺ ង ង Fe ង ម ម Co ង ᝺ ᝺ Ni ង ង ង Cu...3M Tri-M- ite WETORDRY trademark silicon carbide paper, wet with a trichloroethane base coolant, to remove about 0.05 mm in the center. An 0.1

Irradiation testing of high density uranium alloy dispersion fuels

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

Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

1997-10-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U{sub 2}Mo, or U{sub 3}Si{sub 2}. These experiments will be discharged at peak fuel burnups ofmore » 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.« less

Microstructure of RERTR DU-alloys irradiated with krypton ions up to 100 dpa

NASA Astrophysics Data System (ADS)

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

2011-04-01

The radiation stability of the interaction product formed at the fuel-matrix interface of research reactor dispersion fuels, under fission-product bombardment, has a strong impact on fuel performance. Three depleted uranium alloys were cast that consisted of the following five phases to be investigated: U(Si, Al) 3, (U, Mo)(Si, Al) 3, UMo 2Al 20, U 6Mo 4Al 43, and UAl 4. Irradiation of transmission electron microscopy (TEM) disc samples with 500-keV Kr ions at 200 °C to doses up to ˜100 displacements per atom (dpa) were conducted using a 300-keV electron microscope equipped with an ion accelerator. TEM results show that the U(Si, Al) 3 and UAl 4 phases remain crystalline at 100 dpa without forming voids. The (U, Mo)(Si, Al) 3 and UMo 2Al 20 phases become amorphous at 1 and ˜2 dpa, respectively, and show no evidence of voids at 100 dpa. The U 6Mo 4Al 43 phase goes to amorphous at less than 1 dpa and reveals high density voids at 100 dpa.

Effect of molybdenum ion implantation of the pitting corrosion of depleted uranium - 0.75 titanium alloy. (Reannouncement with new availability information). Final report

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

Lei, K.S.; Chang, F.; Levy, M.

1993-07-01

Pitting corrosion of molybdenum-ion-implanted, depleted uranium -0 75 Ti (DU -0 75 Ti) has been studied electrochemically in acidic, neutral, and alkaline solutions containing sodium chloride, and the results have been compared to those of the unimplanted DU -0 75 Ti. The data show that Mo implantation shifts the pitting potential of DU -0 75 Ti in the noble direction in acidic and alkaline solutions. In neutral 50 ppm Cl- solution, however, there is no beneficial effect of Mo implantation. Auger analysis studies show that before exposure to the solutions, all the molybdenum is in the oxide, which is approximatelymore » l000 A thick. After electrochemical scans in the acidic and alkaline chloride solutions, most of the Mo disappears from the oxide. However, no decrease in Mo concentration is found after exposure in neutral chloride solution. It is proposed that the implanted molybdenum dissolves in the acidic and alkaline solutions and forms simple or complex molybdates that inhibit pitting corrosion. The implanted molybdenum does not dissolve in the neutral chloride solution and inhibition does not occur.« less

Use of the Hugoniot elastic limit in laser shockwave experiments to relate velocity measurements

NASA Astrophysics Data System (ADS)

Smith, James A.; Lacy, Jeffrey M.; Lévesque, Daniel; Monchalin, Jean-Pierre; Lord, Martin

2016-02-01

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. This fuel-cladding interface qualification will ensure the survivability of the fuel plates in the harsh reactor environment even under abnormal operating conditions. One of the concerns of the project is the difficulty of calibrating and standardizing the laser shock technique. An analytical study under development and experimental testing supports the hypothesis that the Hugoniot Elastic Limit (HEL) in materials can be a robust and simple benchmark to compare stresses generated by different laser shock systems.

Quantitative ion beam analysis of M-C-O systems: application to an oxidized uranium carbide sample

NASA Astrophysics Data System (ADS)

Martin, G.; Raveu, G.; Garcia, P.; Carlot, G.; Khodja, H.; Vickridge, I.; Barthe, M. F.; Sauvage, T.

2014-04-01

A large variety of materials contain both carbon and oxygen atoms, in particular oxidized carbides, carbon alloys (as ZrC, UC, steels, etc.), and oxycarbide compounds (SiCO glasses, TiCO, etc.). Here a new ion beam analysis methodology is described which enables quantification of elemental composition and oxygen concentration profile over a few microns. It is based on two procedures. The first, relative to the experimental configuration relies on a specific detection setup which is original in that it enables the separation of the carbon and oxygen NRA signals. The second concerns the data analysis procedure i.e. the method for deriving the elemental composition from the particle energy spectrum. It is a generic algorithm and is here successfully applied to characterize an oxidized uranium carbide sample, developed as a potential fuel for generation IV nuclear reactors. Furthermore, a micro-beam was used to simultaneously determine the local elemental composition and oxygen concentration profiles over the first microns below the sample surface. This method is adapted to the determination of the composition of M?C?O? compounds with a sensitivity on elemental atomic concentrations around 1000 ppm.

Theoretical Estimate of Maximum Possible Nuclear Explosion

DOE R&D Accomplishments Database

Bethe, H. A.

1950-01-31

The maximum nuclear accident which could occur in a Na-cooled, Be moderated, Pu and power producing reactor is estimated theoretically. (T.R.H.) 2O82 Results of nuclear calculations for a variety of compositions of fast, heterogeneous, sodium-cooled, U-235-fueled, plutonium- and power-producing reactors are reported. Core compositions typical of plate-, pin-, or wire-type fuel elements and with uranium as metal, alloy, and oxide were considered. These compositions included atom ratios in the following range: U-23B to U-235 from 2 to 8; sodium to U-235 from 1.5 to 12; iron to U-235 from 5 to 18; and vanadium to U-235 from 11 to 33. Calculations were performed to determine the effect of lead and iron reflectors between the core and blanket. Both natural and depleted uranium were evaluated as the blanket fertile material. Reactors were compared on a basis of conversion ratio, specific power, and the product of both. The calculated results are in general agreement with the experimental results from fast reactor assemblies. An analysis of the effect of new cross-section values as they became available is included. (auth)

Some emerging applications of lasers

NASA Astrophysics Data System (ADS)

Christensen, C. P.

1982-10-01

Applications of lasers in photochemistry, advanced instrumentation, and information storage are discussed. Laser microchemistry offers a number of new methods for altering the morphology of a solid surface with high spatial resolution. Recent experiments in material deposition, material removal, and alloying and doping are reviewed. A basic optical disk storage system is described and the problems faced by this application are discussed, in particular those pertaining to recording media. An advanced erasable system based on the magnetooptic effect is described. Applications of lasers for remote sensing are discussed, including various lidar systems, the use of laser-induced fluorescence for oil spill characterization and uranium exploration, and the use of differential absorption for detection of atmospheric constituents, temperature, and humidity.

A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

NASA Astrophysics Data System (ADS)

Van Kleeck, Melissa A.

The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

Oxidation of U-20 at% Zr alloy in air at 423 1063 K

NASA Astrophysics Data System (ADS)

Matsui, Tsuneo; Yamada, Takanobu; Ikai, Yasushi; Naito, Keiji

1993-01-01

The oxidation behavior of U 0.80Zr 0.20 alloy (two-phase mixture of U and UZr 2 below 878 K and single solid solution above 1008 K) was studied by thermogravimetry in the temperature range from 423 to 1063 K in air. During oxidation in the low temperature region (423-503 K), the sample kept its initial shape (a rectangular rod) and the surface of the sample was covered by a black thin adherent UO2 + x oxide layer. On the other hand, by oxidation in the middle temperature region, the sample broke to several pieces of thin plates and blocks, and fine powder at 643-723 K and entirely to fine powder at 775-878 K, all of which were analyzed to be a mixture of U 3O 8 and ZrO 2. By oxidation in the high temperature region (1008-1063 K) the sample broke to very fine powder, which consisted of U 3O 8 and ZrO 2. Based on the sample shape, the oxide phase identified after oxidation and the slope value of the bilogarithmic plots of the weight gain against time, the oxidation kinetics was analyzed with a paralinear equation in the low temperature region below 503 K and a linear equation in the middle and high temperature regions above 643 K. Oxidation rates of U 0.80Zr 0.20 (two-phase mixture) in the low and middle temperature regions were smaller than those of uranium metal. A discontinuity in the plot of the linear oxidation rate constant versus reciprocal temperature was found to be present between 723 and 838 K, similarly to the case of uranium metal previously reported. The linear rate constants of single-phase solid solution in the high temperature region above 1008 K seemed to be a little smaller than those estimated by the extrapolation of the values in the middle temperature region.

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

NASA Astrophysics Data System (ADS)

Rest, J.; Hofman, G. L.; Kim, Yeon Soo

2009-04-01

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

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

Jian Gan; Brandon Miller; Dennis Keiser

2014-04-01

As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists ofmore » fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.« less

Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

NASA Astrophysics Data System (ADS)

Shi, Xue-Ming; Peng, Xian-Jue

2016-09-01

Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

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

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

Not Available

1991-09-01

In January 1990, the Secretary of Energy approved a plan for the demonstration and deployment of the Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) technology, with the near-term goal to provide the necessary information to make a deployment decision by November 1992. The U-AVLIS process is based on electrostatic extraction of photoionized U-235 atoms from an atomic vapor stream created by electron-beam vaporization of uranium metal alloy. A programmatic document for use in screening DOE sites to locate the U-AVLIS production plant was developed and implemented in two parts (Wolsko et al. 1991). The first part consisted of a seriesmore » of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. These sites were then subjected to a more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. This environmental site description (ESD) provides a detailed description of the ORGDP site and vicinity suitable for use in an environmental impact statement (EIS). The report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during a site visit. The organization of the ESD is as follows. Topics addressed in Sec. 2 include a general site description and the disciplines of geology, water resources, biotic resources, air resources, noise, cultural resources, land use, socioeconomics, and waste management. Identification of any additional data that would be required for an EIS is presented in Sec. 3. Following the site description and additional data requirements, Sec. 4 provides a short, qualitative assessment of potential environmental issues. 37 refs., 20 figs., 18 tabs.« less

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

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

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

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

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated inmore » the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.« less

Computer simulation of structural modifications induced by highly energetic ions in uranium dioxide

NASA Astrophysics Data System (ADS)

Sasajima, Y.; Osada, T.; Ishikawa, N.; Iwase, A.

2013-11-01

The structural modification caused by the high-energy-ion irradiation of single-crystalline uranium dioxide was simulated by the molecular dynamics method. As the initial condition, high kinetic energy was supplied to the individual atoms within a cylindrical region of nanometer-order radius located in the center of the specimen. The potential proposed by Basak et al. [C.B. Basak, A.K. Sengupta, H.S. Kamath, J. Alloys Compd. 360 (2003) 210-216] was utilized to calculate interaction between atoms. The supplied kinetic energy was first spent to change the crystal structure into an amorphous one within a short period of about 0.3 ps, then it dissipated in the specimen. The amorphous track radius Ra was determined as a function of the effective stopping power gSe, i.e., the kinetic energy of atoms per unit length created by ion irradiation (Se: electronic stopping power, g: energy transfer ratio from stopping power to lattice vibration energy). It was found that the relationship between Ra and gSe follows the relation Ra2=aln(gS)+b. Compared to the case of Si and β-cristobalite single crystals, it was harder to produce amorphous track because of the long range interaction between U atoms.

Summary of ORSphere critical and reactor physics measurements

NASA Astrophysics Data System (ADS)

Marshall, Margaret A.; Bess, John D.

2017-09-01

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

Atomistic Simulation of High-Density Uranium Fuels

DOE PAGES

Garcés, Jorge Eduardo; Bozzolo, Guillermo

2011-01-01

We apply an atomistic modeling approach to deal with interfacial phenomena in high-density uranium fuels. The effects of Si, as additive to Al or as U-Mo-particles coating, on the behavior of the Al/U-Mo interface is modeled by using the Bozzolo-Ferrante-Smith (BFS) method for alloys. The basic experimental features characterizing the real system are identified, via simulations and atom-by-atom analysis. These include (1) the trend indicating formation of interfacial compounds, (2) much reduced diffusion of Al into U-Mo solid solution due to the high Si concentration, (3) Si depletion in the Al matrix, (4) an unexpected interaction between Mo and Simore » which inhibits Si diffusion to deeper layers in the U-Mo solid solution, and (5) the minimum amount of Si needed to perform as an effective diffusion barrier. Simulation results related to alternatives to Si dispersed in the Al matrix, such as the use of C coating of U-Mo particles or Zr instead of the Al matrix, are also shown. Recent experimental results confirmed early theoretical proposals, along the lines of the results reported in this work, showing that atomistic computational modeling could become a valuable tool to aid the experimental work in the development of nuclear fuels.« less

Survey of Materials for Fusion Fission Hybrid Reactors Vol 1 Rev. 0

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

Farmer, Joseph Collin

2007-07-03

Materials for fusion-fission hybrid reactors fall into several broad categories, including fuels, blanket and coolant materials, cladding, structural materials, shielding, and in the specific case of inertial-confinement fusion systems, laser and optical materials. This report surveys materials in all categories of materials except for those required for lasers and optics. Preferred collants include two molten salt mixtures known as FLIBE (Li2BeF4) and FLINABE (LiNaBeF4). In the case of homogenous liquid fuels, UF4 can be dissolved in these molten salt mixtures. The transmutation of lithium in this coolant produces very corrosive hydrofluoric acid species (HF and TF), which can rapidly degrademore » structural materials. Broad ranges of high-melting radiation-tolerant structural material have been proposed for fusion-fission reactor structures. These include a wide variety of steels and refractory alloys. Ferritic steels with oxide-dispersion strengthening and graphite have been given particular attention. Refractory metals are found in Groups IVB and VB of the periodic table, and include Nb, Ta, Cr, Mo, and W, as serve as the basis of refractory alloys. Stable high-melting composites and amorphous metals may also be useful. Since amorphous metals have no lattice structure, neutron bombardment cannot dislodge atoms from lattice sites, and the materials would be immune from this specific mode of degradation. The free energy of formation of fluorides of the alloying elements found in steels and refractory alloys can be used to determine the relative stability of these materials in molten salts. The reduction of lithium transmutation products (H + and T +) drives the electrochemical corrosion process, and liberates aggressive fluoride ions that pair with ions formed from dissolved structural materials. Corrosion can be suppressed through the use of metallic Be and Li, though the molten salt becomes laden with colloidal suspensions of Be and Li corrosion products in the process. Alternatively, imposed currents and other high-temperature cathodic protection systems are envisioned for protection of the structural materials. This novel concept could prove to be enabling technology for such high-temperature molten-salt reactors. The use of UF 4 as a liquid-phase homogenous fuel is also complicated by redox control. For example, the oxidation of tetravalent uranium to hexavalent uranium could result in the formation of volatile UF 6. This too could be controlled through electrochemically manipulated oxidation and reduction reactions. In situ studies of pertinent electrochemical reactions in the molten salts are proposed, and are relevant to both the corrosive attack of structural materials, as well as the volatilization of fuel. Some consideration is given to the potential advantages of gravity fed falling-film blankets. Such systems may be easier to control than vortex systems, but would require that cylindrical reaction vessels be oriented with the centerline normal to the gravitational field.« less

Characterization Of Nuclear Materials Using Time-Of-Flight ICP-MS

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

Buerger, Stefan; Riciputi, Lee R; Bostick, Debra A

2006-01-01

The investigation of illicit trafficking of nuclear materials, nuclear safeguards analysis, and non-proliferation control requires sensitive and isotope-selective detection methods to gain crucial nuclear forensic information like isotope 'fingerprints' and multi-element signatures. The advantage of time-of-flight (TOF) mass spectrometry - quasi-simultaneous multi-mass analysis - combined with an inductively coupled plasma (ICP) ion source provides an analytical instrument with multi-element and multi-isotope capability and good detection limits. A TOF-ICP-MS system thus appears to be an advantageous choice for the investigation and characterization of nuclear materials. We present here results using a GBC OptiMass 8000 time-of-flight ICP-MS for the isotope screening ofmore » solid samples by laser ablation and the multi-element determination of impurities in uranium ore concentrates using matrix matched standards. A laser ablation system (New Wave Research, UP 213) coupled to the TOF-ICP-MS instrument has been used to optimize the system for analysis of non-radioactive metal samples of natural isotopic composition for a variety of elements including Cu, Sr, Zr, Mo, Cd, In, Ba, Ta, W, Re, Pt, and Pb in pure metals, alloys, and glasses to explore precision, accuracy, and detection limits. Similar methods were then applied to measure uranium. When the laser system is optimized, no mass bias correction is required. Precision and accuracy for the determination of the isotopic composition is typically 1 - 3% for elemental concentrations of as little as 50 ppm in the matrix, with no requirement for sample preparation. The laser ablation precision and accuracy are within ~10x of the instrumental limits for liquid analysis (0.1%). We have investigated the capabilities of the TOF-ICP-MS for the analysis of impurities in uranium matrices. Matrix matching has been used to develop calibration curves for a range of impurities (alkaline, earth-alkaline, transition metals, and rare earth elements). These calibration curves have been used to measure impurities in a number of uranium samples. The results from the TOF-ICP-MS will be compared with other mass spectrometric methods.« less

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.

Embedded Weapons-Grade Tungsten Alloy Shrapnel Rapidly Induces Metastatic High-Grade Rhabdomyosarcomas in F344 Rats

PubMed Central

Kalinich, John F.; Emond, Christy A.; Dalton, Thomas K.; Mog, Steven R.; Coleman, Gary D.; Kordell, Jessica E.; Miller, Alexandra C.; McClain, David E.

2005-01-01

Continuing concern regarding the potential health and environmental effects of depleted uranium and lead has resulted in many countries adding tungsten alloy (WA)-based munitions to their battlefield arsenals as replacements for these metals. Because the alloys used in many munitions are relatively recent additions to the list of militarily relevant metals, very little is known about the health effects of these metals after internalization as embedded shrapnel. Previous work in this laboratory developed a rodent model system that mimicked shrapnel loads seen in wounded personnel from the 1991 Persian Gulf War. In the present study, we used that system and male F344 rats, implanted intramuscularly with pellets (1 mm × 2 mm cylinders) of weapons-grade WA, to simulate shrapnel wounds. Rats were implanted with 4 (low dose) or 20 pellets (high dose) of WA. Tantalum (20 pellets) and nickel (20 pellets) served as negative and positive controls, respectively. The high-dose WA-implanted rats (n = 46) developed extremely aggressive tumors surrounding the pellets within 4–5 months after implantation. The low-dose WA-implanted rats (n = 46) and nickel-implanted rats (n = 36) also developed tumors surrounding the pellets but at a slower rate. Rats implanted with tantalum (n = 46), an inert control metal, did not develop tumors. Tumor yield was 100% in both the low- and high-dose WA groups. The tumors, characterized as high-grade pleomorphic rhabdomyosarcomas by histopathology and immunohistochemical examination, rapidly metastasized to the lung and necessitated euthanasia of the animal. Significant hematologic changes, indicative of polycythemia, were also observed in the high-dose WA-implanted rats. These changes were apparent as early as 1 month postimplantation in the high-dose WA rats, well before any overt signs of tumor development. These results point out the need for further studies investigating the health effects of tungsten and tungsten-based alloys. PMID:15929896

Posttest examination results of recent treat tests on metal fuel

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

Holland, J.W.; Wright, A.E.; Bauer, T.H.

A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity.more » Eutectic penetration and failure of the cladding were also examined in the failed pins.« less

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

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

Heidet, F.; Kim, T.; Grandy, C.

2012-07-30

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium ismore » more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.« less

Development of Nitride Coating Using Atomic Layer Deposition for Low-Enriched Uranium Fuel Powder

NASA Astrophysics Data System (ADS)

Bhattacharya, Sumit

High-performance research reactors require fuel that operates at high specific power and can withstand high fission density, but at relatively low temperatures. The design of the research reactor fuels is done for efficient heat emission, and consists of assemblies of thin-plates cladding made from aluminum alloy. The low-enriched fuels (LEU) were developed for replacing high-enriched fuels (HEU) for these reactors necessitates a significantly increased uranium density in the fuel to counterbalance the decrease in enrichment. One of the most promising new fuel candidate is U-Mo alloy, in a U-Mo/Al dispersion fuel form, due to its high uranium loading as well as excellent irradiation resistance performance, is being developed extensively to convert from HEU fuel to LEU fuel for high-performance research reactors. However, the formation of an interaction layer (IL) between U-Mo particles and the Al matrix, and the associated pore formation, under high heat flux and high burnup conditions, degrade the irradiation performance of the U-Mo/Al dispersion fuel. From the recent tests results accumulated from the surface engineering of low enriched uranium fuel (SELENIUM) and MIR reactor displayed that a surface barrier coating like physical vapor deposited (PVD) zirconium nitride (ZrN) can significantly reduce the interaction layer. The barrier coating performed well at low burn up but above a fluence rate of 5x 1021 ions/cm2 the swelling reappeared due to formation interaction layer. With this result in mind the objective of this research was to develop an ultrathin ZrN coating over particulate uranium-molybdenum nuclear fuel using a modified savannah 200 atomic layer deposition (ALD) system. This is done in support of the US Department of Energy's (DOE) effort to slow down the interaction at fluence rate and reach higher burn up for high power research reactor. The low-pressure Savannah 200 ALD system is modified to be designed as a batch powder coating system using the metal organic chemical precursors tetrakis dimethylamido zirconium (TDMAZr) and ammonia( NH3) for succesful deposition of ZrN coating. Nitrogen (N2) gas carried the chemicals to a hot wall reactor maintained at a temperature range of 235 to 245 °C. The ALD system design evolved over the course of this research as the process variables were steadily improved. The conditions found deemed for attaining best coating were at a temperature of 245 °C, with pulse time of 0.8 seconds for TDMAZr and 0.1 seconds for NH3 along with 15 seconds of purge time in-between each cycle. The ALD system was successful in making 1-micrometer (um) ZrN with low levels of chemical impurities over U-Mo powder batches. The deposited coatings were characterized using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), electron energy loss spectroscopy (EELS) and Transmission electron microscope (TEM). This document describes the establishment of the Savannah 200 ALD system, precursor surface reaction procedures and finally the nature of the coating achieved, including characterization of the coating at the different stages of deposition. It was found that an interlayer of alumina in between ZrN and the U-Mo surface was required to reduce the residual stress generated during the ALD procedure. The alumina not only removed the risk of cracking and spallation of the ZrN coating but also provided adequate strength for the barrier layer to withstand the fuel plate rolling conditions. The ZrN coating was nano crystalline in nature, with grain size varying from 5-10 nm, the deposited layer was found to be dense consisting of a layered structure. The coating could retain its crystallinity and maintain its phase when irradiated with 1 MeV single charged ion Kr to produce a damage of 10 displacement per atom (DPA) at intermediate voltage electron microscopy (IVEM).

Fermi energy 5f spectral weight variation in uranium alloys

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

Denlinger, J.D.; Clack, J.; Allen, J.W.

1997-04-01

Uranium materials display a wide range of thermal, electrical and magnetic properties, often exotic. For more than a decade there have been efforts to use photoemission spectroscopy to develop a systematic and unified understanding of the 5f electron states giving rise to this behavior. These efforts have been hampered by a paucity of systems where changes in transport properties are accompanied by substantial spectral changes, so as to allow an attempt to correlate the two kinds of properties within some model. The authors have made resonant photoemission measurements to extract the 5f spectral weight in three systems which show varyingmore » degrees of promise of permitting such an attempt, Y{sub 1{minus}x}U{sub x}Pd{sub 3}, U(Pd{sub x}Pt{sub 1{minus}x}){sub 3} and U(Pd{sub x}Cu{sub 1{minus}x}){sub 5}. They have also measured U 4f core level spectra. The 4f spectra can be modeled with some success by the impurity Anderson model (IAM), and the 5f spectra are currently being analyzed in that framework. The IAM characterizes the 5f-electrons of a single site by an f binding energy {epsilon}{sub f}, an f Coulomb interaction and a hybridization V to conduction electrons. Latent in the model are the phenomena of 5f mixed valence and the Kondo effect.« less

Summary of ORSphere Critical and Reactor Physics Measurements

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

Marshall, Margaret A.; Bess, John D.

In the early 1970s Dr. John T. Mihalczo (team leader), J. J. Lynn, and J. R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVAmore » I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.« less

SAFARI-1: Achieving conversion to LEU - A local challenge

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

Piani, C.S.B.

2008-07-15

Two years have passed since the South African Department of Minerals and Energy authorised the conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the South African Research Reactor (SAFARI-1) and the associated fuel manufacturing at Pelindaba. The scheduling, as originally proposed, allowed approximately three years for the full conversion of the reactor, anticipating simultaneous manufacturing ability from the fuel production plant. Due to technical difficulties experienced in the conversion of the local manufacturing plant from HEU (UAl alloy) to LEU (U Silicide) and the uncertainty as to costing and scheduling of such an achievement, the conversionmore » of SAFARI-1 based on local supply has been allocated a lower priority. The acquisition in mid-2006 of 2 LEU silicide elements of SA design, manufactured by AREVA- CERCA and irradiated as test elements in SAFARI-1 to burn-ups of {approx}65% each; was successfully accomplished within 9 cycles of irradiation each. Furthermore, four 'Hybrid' elements (AREVA-CERCA plates assembled locally at Pelindaba) are ready for irradiation and have received regulatory authorisation to load. This will enable the SAFARI-1 conversion program to continue systematically according to an agreed schedule. This paper will trace the developments of the above and reflect the current status and the rescheduled conversion phases of the reactor according to latest expectations. (author)« less

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

Margaret A. Marshall

In the early 1970’s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950’s (HEU-MET-FAST-001). The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared withmore » the GODIVA I experiments. “The very accurate description of this sphere, as assembled, establishes it as an ideal benchmark for calculational methods and cross-section data files.” (Reference 1) While performing the ORSphere experiments care was taken to accurately document component dimensions (±0. 0001 in. for non-spherical parts), masses (±0.01 g), and material data The experiment was also set up to minimize the amount of structural material in the sphere proximity. A three part sphere was initially assembled with an average radius of 3.4665 in. and was then machined down to an average radius of 3.4420 in. (3.4425 in. nominal). These two spherical configurations were evaluated and judged to be acceptable benchmark experiments; however, the two experiments are highly correlated.« less

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

none

This report is the first quarterly report under Contract AT(30-1)-2528 for the decontamination and recovery of precious metals from contaminated scrap. This program encompasses the recovery of gold, platinum, iridium, rhodium and palladium from scrap and contaminated laboratory ware, equipment, etc. which may be contaminated by alpha emitters (i.e., uranium, plutonium and possibly polonium or combinations of these), beta-gamma emitters such as fission products or all three sources of radiation. In addition the scrap can be chemically contaminated by base metals, graphite or other contaminants. Osmium, ruthenium or silver are not to be recovered. Since the scrap can exist inmore » such a great variety of forms such as platinum-rhodium laboratory crucibles, graphite crucibles containing platinum-base metal alloys derived from vacuum fusion baths, insulated thermocouple wires, reactor liners, porous platinum filters, alloys such as Baker alloy 413 (60 Au, 25 Pd, 15 Pt) it may be necessary to devise specific methods for a given type of scrap. Nevertheless the basic chemistry is the same and information and methods applicable to single systems will apply directly to the more complicated ones. Accordingly the relatively simple systems would be investigated first and the more complicated ones later in the program. An essential and major contribution to the success of the program is to devise or adapt accurate and precise analytical chemical techniques for scrap identification, process control, purity of the final products and accountability control. Accordingly, the emphasis on chemical analysis will be proportionally greater during the early stages of the program. The residual activity level to be attained is presently stated at twice the standard deviation obtained in low background counters (alpha {approx}0.1 c/m, beta {approx} 5 c/m) of virgin platinum.« less

Analyzing the impact of reactive transport on the repository performance of TRISO fuel

NASA Astrophysics Data System (ADS)

Schmidt, Gregory

One of the largest determiners of the amount of electricity generated by current nuclear reactors is the efficiency of the thermodynamic cycle used for power generation. Current light water reactors (LWR) have an efficiency of 35% or less for the conversion of heat energy generated by the reactor to electrical energy. If this efficiency could be improved, more power could be generated from equivalent volumes of nuclear fuel. One method of improving this efficiency is to use a coolant flow that operates at a much higher temperature for electricity production. A reactor design that is currently proposed to take advantage of this efficiency is a graphite-moderated, helium-cooled reactor known as a High Temperature Gas Reactor (HTGR). There are significant differences between current LWR's and the proposed HTGR's but most especially in the composition of the nuclear fuel. For LWR's, the fuel elements consist of pellets of uranium dioxide or plutonium dioxide that are placed in long tubes made of zirconium metal alloys. For HTGR's, the fuel, known as TRISO (TRIstructural-ISOtropic) fuel, consists of an inner sphere of fissile material, a layer of dense pyrolytic carbon (PyC), a ceramic layer of silicon carbide (SiC) and a final dense outer layer of PyC. These TRISO particles are then compacted with graphite into fuel rods that are then placed in channels in graphite blocks. The blocks are then arranged in an annular fashion to form a reactor core. However, this new fuel form has unanswered questions on the environmental post-burn-up behavior. The key question for current once-through fuel operations is how these large irradiated graphite blocks with spent fuel inside will behave in a repository environment. Data in the literature to answer this question is lacking, but nevertheless this is an important question that must be answered before wide-spread adoption of HTGR's could be considered. This research has focused on answering the question of how the large quantity of graphite surrounding the spent HTGR fuel will impact the release of aqueous uranium from the TRISO fuel. In order to answer this question, the sorption and partitioning behavior of uranium to graphite under a variety of conditions was investigated. Key systematic variables that were analyzed include solution pH, dissolved carbonate concentration, uranium metal concentration and ionic strength. The kinetics and desorption characteristics of uranium/graphite partitioning were studied as well. The graphite used in these experiments was also characterized by a variety of techniques and conclusions are drawn about the relevant surface chemistry of graphite. This data was then used to generate a model for the reactive transport of uranium in a graphite matrix. This model was implemented with the software code CXTFIT and validated through the use of column studies mirroring the predicted system.

Dosimetry characterization of the Godiva Reactor under burst conditions

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

Hickman, D. P.; Heinrichs, D. P.; Hudson, R.

2017-06-22

A series of sixteen (16) burst irradiations were performed in May 2014, fifteen of which were part of an international collaboration to characterize the Godiva IV fast burst reactor at the National Criticality Experiments Research Center (NCERC). Godiva IV is a bare cylindrical assembly of approximately 65 kg of highly enriched uranium fuel (93.2% 235U metal alloyed with 1.5% molybdenum for strength) and is designed to perform controlled prompt critical excursions (Myers 2010, Goda 2013). Twelve of the irradiations were dedicated to neutron spectral measurements using a Bonner multiple sphere spectrometer. Three irradiations, with core temperature increases of 71.1°C, 136.9°C,more » and 229.9°C, were performed for generating comparative fluence data, establishing corrections for varying heights, testing linearity with burst temperature, and establishing gamma dose characteristics.« less

Corrosion Assessment of Candidate Materials for the SHINE Subcritical Assembly Vessel and Components FY14 Report

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

Pawel, Steven J.

2014-10-01

Laboratory corrosion testing of candidate alloys—including Zr-4 and Zr-2.5Nb representing the target solution vessel, and 316L, 2304, 304L, and 17-4 PH stainless steels representing process piping and balance-of-plant components—was performed in support of the proposed SHINE process to produce 99Mo from low-enriched uranium. The test solutions used depleted uranyl sulfate in various concentrations and incorporated a range of temperatures, excess sulfuric acid concentrations, nitric acid additions (to simulate radiolysis product generation), and iodine additions. Testing involved static immersion of coupons in solution and in the vapor above the solution, and was extended to include planned-interval tests to examine details associatedmore » with stainless steel corrosion in environments containing iodine species. A large number of galvanic tests featuring couples between a stainless steel and a zirconium-based alloy were performed, and limited vibratory horn testing was incorporated to explore potential erosion/corrosion features of compatibility. In all cases, corrosion of the zirconium alloys was observed to be minimal, with corrosion rates based on weight loss calculated to be less than 0.1 mil/year with no change in surface roughness. The resulting passive film appeared to be ZrO2 with variations in thickness that influence apparent coloration (toward light brown for thicker films). Galvanic coupling with various stainless steels in selected exposures had no discernable effect on appearance, surface roughness, or corrosion rate. Erosion/corrosion behavior was the same for zirconium alloys in uranyl sulfate solutions and in sodium sulfate solutions adjusted to a similar pH, suggesting there was no negative effect of uranium resulting from fluid dynamic conditions aggressive to the passive film. Corrosion of the candidate stainless steels was similarly modest across the entire range of exposures. However, some sensitivity to corrosion of the stainless steels was observed in solutions with 50 wppm iodine (the actual SHINE process expects 0.1–1 wppm) with the highest corrosion rates (up to ~6 mil/year) observed on specimens exposed in the vapor phase. Lower concentrations of iodine species (5 or 28 wppm) proved much less corrosive, and the planned-interval data indicated that metal corrodibility decreased with time for all immersed exposures and, with one minor exception, all vapor exposures. Little change in susceptibility to corrosion was observed as a result of nitric acid additions to the test environment (simulating radiolysis products). The trend toward reduced corrosion (immersion and vapor phase) with decreasing iodine concentration suggests that, at the expected conditions in the SHINE process, it is unlikely that iodine species will generate a general corrosion concern for the candidate stainless steels.« less

Determining Coolant Flow Rate Distribution In The Fuel-Modified TRIGA Plate Reactor

NASA Astrophysics Data System (ADS)

Puji Hastuti, Endiah; Widodo, Surip; Darwis Isnaini, M.; Geni Rina, S.; Syaiful, B.

2018-02-01

TRIGA 2000 reactor in Bandung is planned to have the fuel element replaced, from cylindrical uranium and zirconium-hydride (U-ZrH) alloy to U3Si2-Al plate type of low enriched uranium of 19.75% with uranium density of 2.96 gU/cm3, while the reactor power is maintained at 2 MW. This change is planned to anticipate the discontinuity of TRIGA fuel element production. The selection of this plate-type fuel element is supported by the fact that such fuel type has been produced in Indonesia and used in MPR-30 safely since 2000. The core configuration of plate-type-fuelled TRIGA reactor requires coolant flow rate through each fuel element channel in order to meet its safety function. This paper is aimed to describe the results of coolant flow rate distribution in the TRIGA core that meets the safety function at normal operation condition, physical test, shutdown, and at initial event of loss of coolant flow due power supply interruption. The design analysis to determine coolant flow rate in this paper employs CAUDVAP and COOLODN computation code. The designed coolant flow rate that meets the safety criteria of departure from nucleate boiling ratio (DNBR), onset of flow instability ratio (OFIR), and ΔΤ onset of nucleate boiling (ONB), indicates that the minimum flow rate required to cool the plate-type fuelled TRIGA core at 2 MW is 80 kg/s. Therefore, it can be concluded that the operating limitation condition (OLC) for the minimum flow rate is 80 kg/s; the 72 kg/s is to cool the active core; while the minimum flow rate for coolant flow rate drop is limited to 68 kg/s with the coolant inlet temperature 35°C. This thermohydraulic design also provides cooling for 4 positions irradiation position (IP) utilization and 1 central irradiation position (CIP) with end fitting inner diameter (ID) of 10 mm and 20 mm, respectively.

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

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

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-19

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritiummore » allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.« less

Depleted uranium dioxide melting in cold crucible melter and production of granules from the melt for use in casks for spent nuclear fuel and radioactive wastes

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

Gotovchikov, V.T.; Seredenko, V.A.; Shatalov, V.V.

2007-07-01

This paper describes the results of a joint research program between the Russian Research Institute of Chemical Technology and Oak Ridge National Laboratory in the United States to develop new radiation shielding materials for use in the construction of casks for spent nuclear fuel (SNF) and radioactive wastes. Research and development is underway to develop SNF storage, transport, and disposal casks using shielding made with two new depleted uranium dioxide (DUO{sub 2}) materials: a DUO{sub 2}-steel cermet, and, DUCRETE with DUAGG (DUO{sub 2} aggregate). Melting the DUO{sub 2} and allowing it to freeze will produce a near 100% theoretical densitymore » product and assures that the product produces no volatile materials upon subsequent heating. Induction cold-crucible melters (ICCM) are being developed for this specific application. An ICCM is, potentially, a high throughput low-cost process. Schematics of a pilot facility were developed for the production of molten DUO{sub 2} from DU{sub 3}O{sub 8} to produce granules <1 mm in diameter in a continuous mode of operation. Thermodynamic analysis was conducted for uranium-oxygen system in the temperature range from 300 to 4000 K in various gas mediums. Temperature limits of stability for various uranium oxides were determined. Experiments on melting DUO{sub 2} were carried out in a high frequency ICCM in a cold crucible with a 120 mm in diameter. The microstructure of molten DUO{sub 2} was studied and lattice parameters were determined. It was experimentally proved, and validated by X-ray analysis, that an opportunity exists to produce molten DUO{sub 2} from mixed oxides (primarily DU{sub 3}O{sub 8}) by reduction melting in ICCM. This will allow using DU{sub 3}O{sub 8} directly to make DUO{sub 2}-a separate unit operation to produce UO{sub 2} feed material is not needed. Experiments were conducted concerning the addition of alloying components, gadolinium et al. oxides, into the DUO{sub 2} melt while in the crucible. These additives improve neutron and gamma radiation shielding and operation properties of the final solids. Cermet samples of 50 wt % DUO{sub 2} were produced. (authors)« less

31 CFR 540.317 - Uranium feed; natural uranium feed.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Uranium feed; natural uranium feed...) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The term uranium feed or natural uranium feed means natural uranium in the form of UF6 suitable for uranium...

31 CFR 540.317 - Uranium feed; natural uranium feed.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium feed; natural uranium feed...) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The term uranium feed or natural uranium feed means natural uranium in the form of UF6 suitable for uranium...

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.

Evolution of interphase and intergranular strain in zirconium-niobium alloys during deformation at room temperature

NASA Astrophysics Data System (ADS)

Cai, Song

Zr-2.5Nb is currently used for pressure tubes in the CANDU (CANada Deuterium Uranium) reactor. A complete understanding of the deformation mechanism of Zr-2.5Nb is important if we are to accurately predict the in-reactor performance of pressure tubes and guarantee normal operation of the reactors. This thesis is a first step in gaining such an understanding; the deformation mechanism of ZrNb alloys at room temperature has been evaluated through studying the effect of texture and microstructure on deformation. In-situ neutron diffraction was used to monitor the evolution of the lattice strain of individual grain families along both the loading and Poisson's directions and to track the development of interphase and intergranular strains during deformation. The following experiments were carried out with data interpreted using elasto-plastic modeling techniques: (1) Compression tests of a 100%betaZr material at room temperature. (2) Tension and compression tests of hot rolled Zr-2.5Nb plate material. (3) Compression of annealed Zr-2.5Nb. (4) Cyclic loading of the hot rolled Zr-2.5Nb. (5) Compression tests of ZrNb alloys with different Nb and oxygen contents. The experimental results were interpreted using a combination of finite element (FE) and elasto-plastic self-consistent (EPSC) models. The phase properties and phase interactions well represented by the FE model, the EPSC model successfully captured the evolution of intergranular constraint during deformation and provided reasonable estimates of the critical resolved shear stress and hardening parameters of different slip systems under different conditions. The consistency of the material parameters obtained by the EPSC model allows the deformation mechanism at room temperature and the effect of textures and microstructures of ZrNb alloys to be understood. This work provides useful information towards manufacturing of Zr-2.5Nb components and helps in producing ideal microstructures and material properties for pressure tubes. Also it is helpful in guiding the development of new materials for the next generation of nuclear reactors. Furthermore, the large data set obtained from this study can be used in evaluation and improving current and future polycrystalline deformation models.

Dissolution flowsheet for high flux isotope reactor fuel

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

Foster, T.

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3O 8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2. The HFIR fuel cores will be dissolved and the recovered U will be down-blendedmore » into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H 2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H 2 and other permanent gases in the dissolution offgas allowing the development of H 2 generation rate profiles. The H 2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower flammability limit (LFL) for H 2 at a given Hg concentration.« less

Postirradiation analysis of the latest high uranium density miniplate test: RERTR 8.

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

Hofman, G. L.; Kim, Y. S.; Rest, J.

2008-01-01

Results of destructive examination of fuel miniplates irradiated in the RERTR-8 test are discussed. Metallographic features of dispersion fuel containing fuel particles of U-7wt%Mo with 1wt% Ti or 2wt% Zr are analyzed. It is hypothesized that Zr, either as alloy addition or fission product, may have a destabilizing effect on fission gas behavior. The purpose of miniplate test RERTR-8 was to obtain irradiation performance data on monolithic fuel plates fabricated by friction bonding (FB) and isostatic hot pressing (HIP), as well as dispersion fuel plates that contain U-7Mo fuel particles alloyed with small amounts of Zr or Ti (see Fig.more » 1). The results of the monolithic plates destructively examined to date were presented at the 2007 RERTR meeting in Prague. This paper presents the first results on the dispersion plates with Ti and Zr additions to U-7Mo. The effect of Ti and Zr additions to U-7wt%Mo on the extent of fuel-aluminum interdiffusion, although measureable, is small in absolute terms because of the overwhelming effect of the 5% Si addition to the Al matrix. Ti additions to the U-7wt%Mo have no discernable effect on swelling behavior of the fuel. However, there are indications that the addition of Zr may have a destabilizing effect on fission gas behavior at high burnup.« less

Developing a laser shockwave model for characterizing diffusion bonded interfaces

NASA Astrophysics Data System (ADS)

Lacy, Jeffrey M.; Smith, James A.; Rabin, Barry H.

2015-03-01

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However, because the deposition of laser energy into the containment layer on a specimen's surface is intractably complex, the shock wave energy is inferred from the surface velocity measured on the backside of the fuel plate and the depth of the impression left on the surface by the high pressure plasma pulse created by the shock laser. To help quantify the stresses generated at the interfaces, a finite element method (FEM) model is being utilized. This paper will report on initial efforts to develop and validate the model by comparing numerical and experimental results for back surface velocities and front surface depressions in a single aluminum plate representative of the fuel cladding.

Equation of state, phase stability, and phase transformations of uranium-6 wt. % niobium under high pressure and temperature

NASA Astrophysics Data System (ADS)

Zhang, Jianzhong; Vogel, Sven; Brown, Donald; Clausen, Bjorn; Hackenberg, Robert

2018-05-01

In-situ time-of-flight neutron diffraction experiments were conducted on the uranium-niobium alloy with 6 wt. % Nb (U-6Nb) at pressures up to 4.7 GPa and temperatures up to 1073 K. Upon static compression at room temperature, the monoclinic structure of U-6Nb (α″ U-6Nb) remains stable up to the highest experimental pressure. Based on the pressure-volume measurements at room temperature, the least-squares fit using the finite-strain equation of state (EOS) yields an isothermal bulk modulus of B0 = 127 ± 2 GPa for the α″-phase of U-6Nb. The calculated zero-pressure bulk sound speed from this EOS is 2.706 ± 0.022 km/s, which is in good agreement with the linear extrapolation of the previous Hugoniot data above 12 GPa for α″ U-6Nb, indicating that the dynamic response under those shock-loading conditions is consistent with the stabilization of the initial monoclinic phase of U-6Nb. Upon heating at ambient and high pressures, the metastable α″ U-6Nb exhibits complex transformation paths leading to the diffusional phase decomposition, which are sensitive to applied pressure, stress state, and temperature-time path. These findings provide new insight into the behavior of atypical systems such as U-Nb and suggest that the different U-Nb phases are separated by rather small energies and hence highly sensitive to compositional, thermal, and mechanical perturbations.

High-Resolution Characterization of UMo Alloy Microstructure

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

Devaraj, Arun; Kovarik, Libor; Joshi, Vineet V.

2016-11-30

This report highlights the capabilities and procedure for high-resolution characterization of UMo fuels in PNNL. Uranium-molybdenum (UMo) fuel processing steps, from casting to forming final fuel, directly affect the microstructure of the fuel, which in turn dictates the in-reactor performance of the fuel under irradiation. In order to understand the influence of processing on UMo microstructure, microstructure characterization techniques are necessary. Higher-resolution characterization techniques like transmission electron microscopy (TEM) and atom probe tomography (APT) are needed to interrogate the details of the microstructure. The findings from TEM and APT are also directly beneficial for developing predictive multiscale modeling tools thatmore » can predict the microstructure as a function of process parameters. This report provides background on focused-ion-beam–based TEM and APT sample preparation, TEM and APT analysis procedures, and the unique information achievable through such advanced characterization capabilities for UMo fuels, from a fuel fabrication capability viewpoint.« less

Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

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

Krass, A.W.

2005-12-19

This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. Themore » material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.« less

Irradiation effects on thermal properties of LWR hydride fuel

NASA Astrophysics Data System (ADS)

Terrani, Kurt; Balooch, Mehdi; Carpenter, David; Kohse, Gordon; Keiser, Dennis; Meyer, Mitchell; Olander, Donald

2017-04-01

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

IDAHO CHEMICAL PROCESSING PLANT TECHNICAL PROGRESS REPORT FOR APRIL THROUGH JUNE 1958

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

Stevenson, C.E.

1958-11-01

Processing of uranium -aluminum alloy was continued with slight process modifications. Means for recovering rare gases from dissolver off-gas are described. Results of extensive decontamination procedures required to enable entrance to the continuous dissolver cell are also indicated. Pilot plant studies of dissolving aluminum continuously showed that rates of dissolution were decreased by factors of 2 to 4 as the concentration of nitric acid fed was increased from 5.4 to 11N. The rate of aluminum dissolution was found to be proportional to initial area exposed for pieces of different shape. It was found possible to produce a highly basic aluminummore » nitrate solution at a reasonable rate by dissolving to low concentration in dilute acid, followed by evaporation to the desired level. Uranium exchange rate measurements for the TBP extraction process are described. A canned rotor pump under test with graphite bearings operated 6000 hours with nominal wear. Difficulties were experienced in testing a nutating disc pump. Measurements of the potential of zirconium in hydrofluoric acid as a function of pH confirmed the predicted equation. In teflon vessels, zirconium dissolves a little more rapidly in nitric-hydrofluoric acid mixtures than in glass vessels, presumably due to reaction of fluoride with silica. Titunium alloy Types 55A and 75A were found to resist corrosion by certain boiling nitric-hydrochloric acid mixtures. Initial tests have commenced with a NaK-heated 100 liter/hour pilot plant aluminum nitrate calciner to continue process demonstration. In tests in the smaller pilot plant unit, increasing feed spray air ratio was found to increase particle loading in the cyclone effluent. Laboratory studies indicated that a venturi scrubber using dilute nitric acid at 80 C should remove ruthenium effectively from calciner off-gas. In a pilot plant test in which a significant fraction of ruthenium feed was retained by the alumina, substantial absorption of volatilized ruthenium was obtained. Thermal conductivity of alumina near 3000 F was about 0.26 Btu/hr)(ft)( F). In leaching studies, very little strontium or plutonium was removed by water from alumina calcined at 550 C. Dilute nitric acid, however, extracted strontium from this material to the same degree (~ 50 percent) as from material calcined at 400 C. Concentrated basic aluminum nitrate was produced from simulated aluminum nitrate waste by slow hydrolysis with urea followed by evaporation. Aluminum was efficiently extracted from buffered aluminum nitrate solution by acetylacetone and was stripped back into nitric acid. A filterable aluminum phosphate was precipituted from aluminum nitrate solution by urea hydrolysis; the phosphate effectively carried fission products, however. Spectrophotometric methods were developed for macro and micro quantities of uranium, in the presence of high concentrations of other ions, based on tetrapropylammonium nitrate extraction. (For preceding period see ID0-14443.) (auth)« less

Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

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

Robb, Kevin R.

2015-08-01

Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramicmore » microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, FeCrAl would tend to generate heat and hydrogen from oxidation at a slower rate compared to the zirconium-based alloys in use today. The previous study, [2], of the FeCrAl ATF concept during station blackout (SBO) severe accident scenarios in BWRs was based on simulating short term SBO (STSBO), long term SBO (LTSBO), and modified SBO scenarios occurring in a BWR-4 reactor with MARK-I containment. The analysis indicated that FeCrAl had the potential to delay the onset of fuel failure by a few hours depending on the scenario, and it could delay lower head failure by several hours. The analysis demonstrated reduced in-vessel hydrogen production. However, the work was preliminary and was based on limited knowledge of material properties for FeCrAl. Limitations of the MELCOR code were identified for direct use in modeling ATF concepts. This effort used an older version of MELCOR (1.8.5). Since these analyses, the BWR model has been updated for use in MELCOR 1.8.6 [10], and more representative material properties for FeCrAl have been modeled. Sections 2 4 present updated analyses for the FeCrAl ATF concept response during severe accidents in a BWR. The purpose of the study is to estimate the potential gains afforded by the FeCrAl ATF concept during BWR SBO scenarios.« less

40 CFR 421.320 - Applicability: Description of the secondary uranium subcategory.

Code of Federal Regulations, 2012 CFR

2012-07-01

... secondary uranium subcategory. 421.320 Section 421.320 Protection of Environment ENVIRONMENTAL PROTECTION... CATEGORY Secondary Uranium Subcategory § 421.320 Applicability: Description of the secondary uranium... uranium (including depleted uranium) by secondary uranium facilities. ...

40 CFR 421.320 - Applicability: Description of the secondary uranium subcategory.

Code of Federal Regulations, 2011 CFR

2011-07-01

... secondary uranium subcategory. 421.320 Section 421.320 Protection of Environment ENVIRONMENTAL PROTECTION... CATEGORY Secondary Uranium Subcategory § 421.320 Applicability: Description of the secondary uranium... uranium (including depleted uranium) by secondary uranium facilities. ...

40 CFR 421.320 - Applicability: Description of the secondary uranium subcategory.

Code of Federal Regulations, 2013 CFR

2013-07-01

... secondary uranium subcategory. 421.320 Section 421.320 Protection of Environment ENVIRONMENTAL PROTECTION... CATEGORY Secondary Uranium Subcategory § 421.320 Applicability: Description of the secondary uranium... uranium (including depleted uranium) by secondary uranium facilities. ...

40 CFR 421.320 - Applicability: Description of the secondary uranium subcategory.

Code of Federal Regulations, 2014 CFR

2014-07-01

... secondary uranium subcategory. 421.320 Section 421.320 Protection of Environment ENVIRONMENTAL PROTECTION... CATEGORY Secondary Uranium Subcategory § 421.320 Applicability: Description of the secondary uranium... uranium (including depleted uranium) by secondary uranium facilities. ...

40 CFR 421.320 - Applicability: Description of the secondary uranium subcategory.

Code of Federal Regulations, 2010 CFR

2010-07-01

... secondary uranium subcategory. 421.320 Section 421.320 Protection of Environment ENVIRONMENTAL PROTECTION... CATEGORY Secondary Uranium Subcategory § 421.320 Applicability: Description of the secondary uranium... uranium (including depleted uranium) by secondary uranium facilities. ...

Bioremediation of uranium contamination with enzymatic uranium reduction

USGS Publications Warehouse

Lovley, D.R.; Phillips, E.J.P.

1992-01-01

Enzymatic uranium reduction by Desulfovibrio desulfuricans readily removed uranium from solution in a batch system or when D. desulfuricans was separated from the bulk of the uranium-containing water by a semipermeable membrane. Uranium reduction continued at concentrations as high as 24 mM. Of a variety of potentially inhibiting anions and metals evaluated, only high concentrations of copper inhibited uranium reduction. Freeze-dried cells, stored aerobically, reduced uranium as fast as fresh cells. D. desulfuricans reduced uranium in pH 4 and pH 7.4 mine drainage waters and in uraniumcontaining groundwaters from a contaminated Department of Energy site. Enzymatic uranium reduction has several potential advantages over other bioprocessing techniques for uranium removal, the most important of which are as follows: the ability to precipitate uranium that is in the form of a uranyl carbonate complex; high capacity for uranium removal per cell; the formation of a compact, relatively pure, uranium precipitate.

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.

PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

DOEpatents

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

1959-09-22

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

STRIPPING OF URANIUM FROM ORGANIC EXTRACTANTS

DOEpatents

Crouse, D.J. Jr.

1962-09-01

A liquid-liquid extraction method is given for recovering uranium values from uranium-containing solutions. Uranium is removed from a uranium-containing organic solution by contacting said organic solution with an aqueous ammonium carbonate solution substantially saturated in uranium values. A uranium- containing precipitate is thereby formed which is separated from the organic and aqueous phases. Uranium values are recovered from this separated precipitate. (AE C)

Detection of depleted uranium in urine of veterans from the 1991 Gulf War.

PubMed

Gwiazda, R H; Squibb, K; McDiarmid, M; Smith, D

2004-01-01

American soldiers involved in "friendly fire" accidents during the 1991 Gulf War were injured with depleted-uranium-containing fragments or possibly exposed to depleted uranium via other routes such as inhalation, ingestion, and/or wound contamination. To evaluate the presence of depleted uranium in these soldiers eight years later, the uranium concentration and depleted uranium content of urine samples were determined by inductively coupled plasma mass spectrometry in (a) depleted uranium exposed soldiers with embedded shrapnel, (b) depleted uranium exposed soldiers with no shrapnel, and (c) a reference group of deployed soldiers not involved in the friendly fire incidents. Uranium isotopic ratios measured in many urine samples injected directly into the inductively coupled plasma mass spectrometer and analyzed at a mass resolution m/delta m of 300 appeared enriched in 235U with respect to natural abundance (0.72%) due to the presence of an interference of a polyatomic molecule of mass 234.81 amu that was resolved at a mass resolution m/delta m of 4,000. The 235U abundance measured on uranium separated from these urines by anion exchange chromatography was clearly natural or depleted. Urine uranium concentrations of soldiers with shrapnel were higher than those of the two other groups, and 16 out of 17 soldiers with shrapnel had detectable depleted uranium in their urine. In depleted uranium exposed soldiers with no shrapnel, depleted uranium was detected in urine samples of 10 out of 28 soldiers. The median uranium concentration of urines with depleted uranium from soldiers without shrapnel was significantly higher than in urines with no depleted uranium, though substantial overlap in urine uranium concentrations existed between the two groups. Accordingly, assessment of depleted uranium exposure using urine must rely on uranium isotopic analyses, since urine uranium concentration is not an unequivocal indicator of depleted uranium presence in soldiers with no embedded shrapnel.

31 CFR 540.317 - Uranium feed; natural uranium feed.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Uranium feed; natural uranium feed... (Continued) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The...

31 CFR 540.317 - Uranium feed; natural uranium feed.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Uranium feed; natural uranium feed... (Continued) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The...

31 CFR 540.317 - Uranium feed; natural uranium feed.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Uranium feed; natural uranium feed... (Continued) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The...

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.

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.

Diffusive gradient in thin FILMS (DGT) compared with soil solution and labile uranium fraction for predicting uranium bioavailability to ryegrass.

PubMed

Duquène, L; Vandenhove, H; Tack, F; Van Hees, M; Wannijn, J

2010-02-01

The usefulness of uranium concentration in soil solution or recovered by selective extraction as unequivocal bioavailability indices for uranium uptake by plants is still unclear. The aim of the present study was to test if the uranium concentration measured by the diffusive gradient in thin films (DGT) technique is a relevant substitute for plant uranium availability in comparison to uranium concentration in the soil solution or uranium recovered by ammonium acetate. Ryegrass (Lolium perenne L. var. Melvina) is grown in greenhouse on a range of uranium spiked soils. The DGT-recovered uranium concentration (C(DGT)) was correlated with uranium concentration in the soil solution or with uranium recovered by ammonium acetate extraction. Plant uptake was better predicted by the summed soil solution concentrations of UO(2)(2+), uranyl carbonate complexes and UO(2)PO(4)(-). The DGT technique did not provide significant advantages over conventional methods to predict uranium uptake by plants. Copyright 2009 Elsevier Ltd. All rights reserved.

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.

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.

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

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.

Method of preparation of uranium nitride

DOEpatents

Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

2013-07-09

Method for producing terminal uranium nitride complexes comprising providing a suitable starting material comprising uranium; oxidizing the starting material with a suitable oxidant to produce one or more uranium(IV)-azide complexes; and, sufficiently irradiating the uranium(IV)-azide complexes to produce the terminal uranium nitride complexes.

10 CFR 760.1 - Uranium leases on lands controlled by DOE. (Domestic Uranium Program Circular No. 760.1, formerly...

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Uranium leases on lands controlled by DOE. (Domestic Uranium Program Circular No. 760.1, formerly (AEC) Domestic Uranium Program Circular 8, 10 CFR 60.8). 760.1 Section 760.1 Energy DEPARTMENT OF ENERGY DOMESTIC URANIUM PROGRAM § 760.1 Uranium leases on lands...

10 CFR 760.1 - Uranium leases on lands controlled by DOE. (Domestic Uranium Program Circular No. 760.1, formerly...

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Uranium leases on lands controlled by DOE. (Domestic Uranium Program Circular No. 760.1, formerly (AEC) Domestic Uranium Program Circular 8, 10 CFR 60.8). 760.1 Section 760.1 Energy DEPARTMENT OF ENERGY DOMESTIC URANIUM PROGRAM § 760.1 Uranium leases on lands...

10 CFR 760.1 - Uranium leases on lands controlled by DOE. (Domestic Uranium Program Circular No. 760.1, formerly...

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Uranium leases on lands controlled by DOE. (Domestic Uranium Program Circular No. 760.1, formerly (AEC) Domestic Uranium Program Circular 8, 10 CFR 60.8). 760.1 Section 760.1 Energy DEPARTMENT OF ENERGY DOMESTIC URANIUM PROGRAM § 760.1 Uranium leases on lands...

10 CFR 760.1 - Uranium leases on lands controlled by DOE. (Domestic Uranium Program Circular No. 760.1, formerly...

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Uranium leases on lands controlled by DOE. (Domestic Uranium Program Circular No. 760.1, formerly (AEC) Domestic Uranium Program Circular 8, 10 CFR 60.8). 760.1 Section 760.1 Energy DEPARTMENT OF ENERGY DOMESTIC URANIUM PROGRAM § 760.1 Uranium leases on lands...

10 CFR 760.1 - Uranium leases on lands controlled by DOE. (Domestic Uranium Program Circular No. 760.1, formerly...

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Uranium leases on lands controlled by DOE. (Domestic Uranium Program Circular No. 760.1, formerly (AEC) Domestic Uranium Program Circular 8, 10 CFR 60.8). 760.1 Section 760.1 Energy DEPARTMENT OF ENERGY DOMESTIC URANIUM PROGRAM § 760.1 Uranium leases on lands...

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.

31 CFR 540.309 - Natural uranium.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Natural uranium. 540.309 Section 540... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.309 Natural uranium. The term natural uranium means uranium found in...

31 CFR 540.309 - Natural uranium.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Natural uranium. 540.309 Section 540... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.309 Natural uranium. The term natural uranium means uranium found in...

31 CFR 540.309 - Natural uranium.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Natural uranium. 540.309 Section 540... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.309 Natural uranium. The term natural uranium means uranium found in...

Potential annealing treatments for tailoring the starting microstructure of low-enriched U-Mo dispersion fuels to optimize performance during irradiation

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Woolstenhulme, Nicolas E.; Ewh, Ashley

2011-12-01

Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525 °C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475 °C.

Rapid Radiochemical Method for Isotopic Uranium in Building ...

EPA Pesticide Factsheets

Technical Fact Sheet Analysis Purpose: Qualitative analysis Technique: Alpha spectrometry Method Developed for: Uranium-234, uranium-235, and uranium-238 in concrete and brick samples Method Selected for: SAM lists this method for qualitative analysis of uranium-234, uranium-235, and uranium-238 in concrete or brick building materials. Summary of subject analytical method which will be posted to the SAM website to allow access to the method.

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.

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

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

Cummins, Dustin Ray; Vogel, Sven C.; Hollis, Kendall Jon

2016-10-18

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffractionmore » data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to the process and analysis can be made, and neutron diffraction can become a viable and efficient technique for gamma/alpha phase composition determination for nuclear fuels.« less

Developing a laser shockwave model for characterizing diffusion bonded interfaces

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

Lacy, Jeffrey M., E-mail: Jeffrey.Lacy@inl.gov; Smith, James A., E-mail: Jeffrey.Lacy@inl.gov; Rabin, Barry H., E-mail: Jeffrey.Lacy@inl.gov

2015-03-31

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengthsmore » in fresh and irradiated fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However, because the deposition of laser energy into the containment layer on a specimen's surface is intractably complex, the shock wave energy is inferred from the surface velocity measured on the backside of the fuel plate and the depth of the impression left on the surface by the high pressure plasma pulse created by the shock laser. To help quantify the stresses generated at the interfaces, a finite element method (FEM) model is being utilized. This paper will report on initial efforts to develop and validate the model by comparing numerical and experimental results for back surface velocities and front surface depressions in a single aluminum plate representative of the fuel cladding.« less

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

Margaret A. Marshall; John D. Bess; Yevgeniy Rozhikhin

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s (HEU-MET-FAST-001). The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared withmore » the GODIVA I experiments[1]. Part of the experimental series was the measurement of the delayed neutron fraction, ßeff, using time correlation measurements and using the central void reactivity measurement. The time correlations measurements were rejected by the experimenter. The measurements using the central void reactivity measurement yielded a ßeff value of 0.00657, which agrees well with the value measured with GODIVA I (0.0066). This measurement is evaluated, found to be acceptable, and discussed in extensive detail in “ORSphere: Physics Measurements for Bare, HEU(93.2) Metal Sphere”[2]. In order to determine the delayed neutron fraction using the central void reactivity delayed neutron parameters must be used. The experimenter utilized the delayed neutron parameters set forth by Keepin, Wimment, and Zeigler[3]. If the derivation of the ßeff is repeated with different delayed neutron parameters from various modern nuclear data sets the resulting values vary greatly from the expected results.« less

16. VIEW OF THE ENRICHED URANIUM RECOVERY SYSTEM. ENRICHED URANIUM ...

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

16. VIEW OF THE ENRICHED URANIUM RECOVERY SYSTEM. ENRICHED URANIUM RECOVERY PROCESSED RELATIVELY PURE MATERIALS AND SOLUTIONS AND SOLID RESIDUES WITH RELATIVELY LOW URANIUM CONTENT. URANIUM RECOVERY INVOLVED BOTH SLOW AND FAST PROCESSES. (4/4/66) - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

77 FR 51579 - Application for a License To Export High-Enriched Uranium

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-24

... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant.... Complex, July 30, 2012, August Uranium (93.35%). uranium-235 high-enriched 1, 2012, XSNM3726, 11006037. contained in 7.5 uranium in the kilograms uranium. form of broken metal to the Atomic Energy of Canada...

Deposit model for volcanogenic uranium deposits

USGS Publications Warehouse

Breit, George N.; Hall, Susan M.

2011-01-01

The International Atomic Energy Agency's tabulation of volcanogenic uranium deposits lists 100 deposits in 20 countries, with major deposits in Russia, Mongolia, and China. Collectively these deposits are estimated to contain uranium resources of approximately 500,000 tons of uranium, which amounts to 6 percent of the known global resources. Prior to the 1990s, these deposits were considered to be small (less than 10,000 tons of uranium) with relatively low to moderate grades (0.05 to 0.2 weight percent of uranium). Recent availability of information on volcanogenic uranium deposits in Asia highlighted the large resource potential of this deposit type. For example, the Streltsovskoye district in eastern Russia produced more than 100,000 tons of uranium as of 2005; with equivalent resources remaining. Known volcanogenic uranium deposits within the United States are located in Idaho, Nevada, Oregon, and Utah. These deposits produced an estimated total of 800 tons of uranium during mining from the 1950s through the 1970s and have known resources of 30,000 tons of uranium. The most recent estimate of speculative resources proposed an endowment of 200,000 tons of uranium.

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

Siekhaus, W. J.; Teslich, N. E.; Weber, P. K.

Depleted uranium that included carbide inclusions was sputtered with 30-keV gallium ions or 16-kev cesium ions to depths much greater than the ions’ range, i.e. using steady-state sputtering. The recession of both the uranium’s and uranium carbide’s surfaces and the ion corresponding fluences were used to determine the steady-state target sputtering yields of both uranium and uranium carbide, i.e. 6.3 atoms of uranium and 2.4 units of uranium carbide eroded per gallium ion, and 9.9 uranium atoms and 3.65 units of uranium carbide eroded by cesium ions. The steady state surface composition resulting from the simultaneous gallium or cesium implantationmore » and sputter-erosion of uranium and uranium carbide were calculated to be U₈₆Ga₁₄, (UC)₇₀Ga₃₀ and U₈₁Cs₉, (UC)₇₉Cs₂₁, respectively.« less

Uranium hydrogeochemical and stream sediment reconnaissance of the Newcastle NTMS Quadrangle, Wyoming, including concentrations of forty-two additional elements

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

Goff, S.J.; Sandoval, W.F.; Gallimore, D.L.

1980-06-01

Water and sediment samples were collected and each water sample was analyzed for U, and each sediment sample was analyzed for 43 elements, including U and Th. Uranium concentrations in water samples range from below the detection limit of 0.02 ppB to 702.26 ppB and have a median of 1.73 ppB and a mean of 11.76 ppB. Water samples containing high uranium concentrations generally are associated with known uranium mining activity or units known to be uranium bearing. About one-third of the water samples containing high uranium concentrations were collected from locations within the Pumpkin Buttes and Turnercrest-Ross Districts. Nearlymore » half of the water samples containing high uranium concentrations were collected from locations just west of the Monument Hill and Highland Flats-Box Creek Districts. Similar anomalous uranium concentrations in this region have been reported updip from Exxon's Highland uranium deposits. High uranium concentrations were also found associated with the Lance Creek-Old Woman Anticline District. Uranium concentrations in sediment samples range from 1.14 to 220.70 ppM and have a median of 3.37 ppM and a mean of 4.03 ppM. Throughout the major uranium mining districts of the Powder River Basin, sediment samples with high uranium concentrations were collected from dry streams located near wells producing water samples with high uranium concentrations. High uranium concentrations were also found associated with the Lance Creek oil field where uranium mineralization is known in the White River formation. High uranium concentrations were also found in sediment samples in areas where uranium mineralization is not known. These samples are from dry streams in areas underlain by the White River formation, the Niobrara formation, and the Pierre, Carlisle, Belle Fourche, and Mowry shales.« less

Influence of uranium hydride oxidation on uranium metal behaviour

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

Patel, N.; Hambley, D.; Clarke, S.A.

2013-07-01

This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, ifmore » sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)« less

Influence of acidic and alkaline waste solution properties on uranium migration in subsurface sediments.

PubMed

Szecsody, Jim E; Truex, Mike J; Qafoku, Nikolla P; Wellman, Dawn M; Resch, Tom; Zhong, Lirong

2013-08-01

This study shows that acidic and alkaline wastes co-disposed with uranium into subsurface sediments have significant impact on changes in uranium retardation, concentration, and mass during downward migration. For uranium co-disposal with acidic wastes, significant rapid (i.e., hours) carbonate and slow (i.e., 100 s of hours) clay dissolution resulted, releasing significant sediment-associated uranium, but the extent of uranium release and mobility change was controlled by the acid mass added relative to the sediment proton adsorption capacity. Mineral dissolution in acidic solutions (pH2) resulted in a rapid (<10 h) increase in aqueous carbonate (with Ca(2+), Mg(2+)) and phosphate and a slow (100 s of hours) increase in silica, Al(3+), and K(+), likely from 2:1 clay dissolution. Infiltration of uranium with a strong acid resulted in significant shallow uranium mineral dissolution and deeper uranium precipitation (likely as phosphates and carbonates) with downward uranium migration of three times greater mass at a faster velocity relative to uranium infiltration in pH neutral groundwater. In contrast, mineral dissolution in an alkaline environment (pH13) resulted in a rapid (<10h) increase in carbonate, followed by a slow (10 s to 100 s of hours) increase in silica concentration, likely from montmorillonite, muscovite, and kaolinite dissolution. Infiltration of uranium with a strong base resulted in not only uranium-silicate precipitation (presumed Na-boltwoodite) but also desorption of natural uranium on the sediment due to the high ionic strength solution, or 60% greater mass with greater retardation compared with groundwater. Overall, these results show that acidic or alkaline co-contaminant disposal with uranium can result in complex depth- and time-dependent changes in uranium dissolution/precipitation reactions and uranium sorption, which alter the uranium migration mass, concentration, and velocity. Copyright © 2013 Elsevier B.V. All rights reserved.

Uranium induces oxidative stress in lung epithelial cells

PubMed Central

Periyakaruppan, Adaikkappan; Kumar, Felix; Sarkar, Shubhashish; Sharma, Chidananda S.

2009-01-01

Uranium compounds are widely used in the nuclear fuel cycle, antitank weapons, tank armor, and also as a pigment to color ceramics and glass. Effective management of waste uranium compounds is necessary to prevent exposure to avoid adverse health effects on the population. Health risks associated with uranium exposure includes kidney disease and respiratory disorders. In addition, several published results have shown uranium or depleted uranium causes DNA damage, mutagenicity, cancer and neurological defects. In the current study, uranium toxicity was evaluated in rat lung epithelial cells. The study shows uranium induces significant oxidative stress in rat lung epithelial cells followed by concomitant decrease in the antioxidant potential of the cells. Treatment with uranium to rat lung epithelial cells also decreased cell proliferation after 72 h in culture. The decrease in cell proliferation was attributed to loss of total glutathione and superoxide dismutase in the presence of uranium. Thus the results indicate the ineffectiveness of antioxidant system’s response to the oxidative stress induced by uranium in the cells. PMID:17124605

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

PubMed

Meinrath, A; Schneider, P; Meinrath, G

2003-01-01

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

METHOD FOR RECOVERING URANIUM FROM OILS

DOEpatents

Gooch, L.H.

1959-07-14

A method is presented for recovering uranium from hydrocarbon oils, wherein the uranium is principally present as UF/sub 4/. According to the invention, substantially complete removal of the uranium from the hydrocarbon oil may be effected by intimately mixing one part of acetone to about 2 to 12 parts of the hydrocarbon oil containing uranium and separating the resulting cake of uranium from the resulting mixture. The uranium in the cake may be readily recovered by burning to the oxide.

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.

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.

Controlling intake of uranium in the workplace: Applications of biokinetic modeling and occupational monitoring data

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

Leggett, Richard Wayne; Eckerman, Keith F; McGinn, Wilson

2012-01-01

This report provides methods for interpreting and applying occupational uranium monitoring data. The methods are based on current international radiation protection guidance, current information on the chemical toxicity of uranium, and best available biokinetic models for uranium. Emphasis is on air monitoring data and three types of bioassay data: the concentration of uranium in urine; the concentration of uranium in feces; and the externally measured content of uranium in the chest. Primary Reference guidance levels for prevention of chemical effects and limitation of radiation effects are selected based on a review of current scientific data and regulatory principles for settingmore » standards. Generic investigation levels and immediate action levels are then defined in terms of these primary guidance levels. The generic investigation and immediate actions levels are stated in terms of radiation dose and concentration of uranium in the kidneys. These are not directly measurable quantities, but models can be used to relate the generic levels to the concentration of uranium in air, urine, or feces, or the total uranium activity in the chest. Default investigation and immediate action levels for uranium in air, urine, feces, and chest are recommended for situations in which there is little information on the form of uranium taken into the body. Methods are prescribed also for deriving case-specific investigation and immediate action levels for uranium in air, urine, feces, and chest when there is sufficient information on the form of uranium to narrow the range of predictions of accumulation of uranium in the main target organs for uranium: kidneys for chemical effects and lungs for radiological effects. In addition, methods for using the information herein for alternative guidance levels, different from the ones selected for this report, are described.« less

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

Gregoire, D.C.; Goltz, D.M.; Chakrabarti, C.L.

Graphite furnace atomic absorption spectrometry (GFAAS) is an insensitive technique for determination of uranium. Experiments were conducted using electrothermal vaporization inductively coupled plasma mass spectrometry to investigate the atomization and vaporization of atomic and molecular uranium species in the graphite furnace. ETV-ICP-MS signals for uranium were observed at temperatures well below the appearance temperature of uranium atoms suggesting the vaporization of molecular uranium oxide at temperatures below 2000{degrees}C. Examination of individual uranium ETV-ICP-MS signals reveals the vaporization of uranium carbide at temperatures above 2600{degrees}C. Chemical modifiers such as 0.2% HF and 0.1% CHF{sub 3} in the argon carrier gas, weremore » ineffective in preventing the formation of uranium carbide at 2700{degrees}C. Vaporization of uranium from a tungsten surface using tungsten foil inserted into the graphite tube prevented the formation of uranium carbide and eliminated the ETV-ICP-MS signal suppression caused by a sodium chloride matrix.« less

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

DOEpatents

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

1959-02-24

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

SOLVENT EXTRACTION PROCESS FOR URANIUM RECOVERY

DOEpatents

Clark, H.M.; Duffey, D.

1958-06-17

A process is described for extracting uranium from uranium ore, wherein the uranium is substantially free from molybdenum contamination. In a solvent extraction process for recovering uranium, uranium and molybdenum ions are extracted from the ore with ether under high acidity conditions. The ether phase is then stripped with water at a lower controiled acidity, resaturated with salting materials such as sodium nitrate, and reextracted with the separation of the molybdenum from the uranium without interference from other metals that have been previously extracted.

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...

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.

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.

Fate of Uranium in Wetlands: Impact of Drought Followed by Re-flooding

NASA Astrophysics Data System (ADS)

Gilson, E.; Huang, S.; Koster van Groos, P. G.; Scheckel, K.; Peacock, A. D.; Kaplan, D. I.; Jaffe, P. R.

2014-12-01

Uranium contamination in groundwater can be mitigated in anoxic zones by iron-reducing bacteria that reduce soluble U(VI) to insoluble U(IV) and by uranium immobilization through complexation and sorption. Wetlands often link ground and surface-waters, making them strategic systems for potentially limiting migration of uranium contamination. Little is known about how drought periods that result in the drying of wetland soils, and consequent redox changes, affect uranium fate and transport in wetlands. In order to better understand the fate and stability of immobilized uranium in wetland soils, and how dry periods affect the uranium stability, we dosed saturated wetland mesocosms planted with Scirpus acutus with low levels of uranyl-acetate for 5 months before imposing a 9-day drying period followed by a 13-day rewetting period. Concentrations of uranium in mesocosm effluent increased after rewetting, but the cumulative amount of uranium released in the 13 days following the drying constituted less than 1% of the uranium immobilized in the soil during the 5 months prior to the drought. This low level of remobilization suggests that the uranium immobilized in these soils was not primarily bioreduced U(IV), which could have been oxidized to soluble U(VI) during the drought and released in the effluent during the subsequent flood. XANES analyses confirm that most of the uranium immobilized in the mesocosms was U(VI) sorbed to iron oxides. Compared to mesocosms that did not experience drying or rewetting, mesocosms that were sacrificed immediately after drying and after 13 days of rewetting had less uranium in soil near roots and more uranium on root surfaces. Metal-reducing bacteria only dominated the bacterial community after 13 days of rewetting and not immediately after drying, indicating that these bacteria are not responsible for this redistribution of uranium after the drying and rewetting. Results show that short periods of drought conditions in a wetland may impact uranium distribution, but these conditions may not cause large losses of immobilized uranium from the wetland.

The distribution of uranium and thorium in granitic rocks of the basin and range province, Western United States

USGS Publications Warehouse

McNeal, J.M.; Lee, D.E.; Millard, H.T.

1981-01-01

Some secondary uranium deposits are thought to have formed from uranium derived by the weathering of silicic igneous rocks such as granites, rhyolites, and tuffs. A regional geochemical survey was made to determine the distribution of uranium and thorium in granitic rocks of the Basin and Range province in order to evaluate the potential for secondary uranium occurrences in the area. The resulting geochemical maps of uranium, thorium, and the Th:U ratio may be useful in locating target areas for uranium exploration. The granites were sampled according to a five-level, nested, analysis-of-variance design, permitting estimates to be made of the variance due to differences between:(1) two-degree cells; (2) one-degree cells; (3) plutons; (4) samples; and (5) analyses. The cells are areas described in units of degrees of latitude and longitude. The results show that individual plutons tend to differ in uranium and thorium concentrations, but that each pluton tends to be relatively homogeneous. Only small amounts of variance occur at the two degree and the between-analyses levels. The three geochemical maps that were prepared are based on one-degree cell means. The reproducibility of the maps is U > Th ??? Th:U. These geochemical maps may be used in three methods of locating target areas for uranium exploration. The first method uses the concept that plutons containing the greatest amounts of uranium may supply the greatest amounts of uranium for the formation of secondary uranium occurrences. The second method is to examine areas with high thorium contents, because thorium and uranium are initially highly correlated but much uranium could be lost by weathering. The third method is to locate areas in which the plutons have particularly high Th:U ratios. Because uranium, but not thorium, is leached by chemical weathering, high Th:U ratios suggest a possible loss of uranium and possibly a greater potential for secondary uranium occurrences to be found in the area. ?? 1981.

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.

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.

Uranium concentrations in groundwater, northeastern Washington

USGS Publications Warehouse

Kahle, Sue C.; Welch, Wendy B.; Tecca, Alison E.; Eliason, Devin M.

2018-04-18

A study of uranium in groundwater in northeastern Washington was conducted to make a preliminary assessment of naturally occurring uranium in groundwater relying on existing information and limited reconnaissance sampling. Naturally occurring uranium is associated with granitic and metasedimentary rocks, as well as younger sedimentary deposits, that occur in this region. The occurrence and distribution of uranium in groundwater is poorly understood. U.S. Environmental Protection Agency (EPA) regulates uranium in Group A community water systems at a maximum contaminant level (MCL) of 30 μg/L in order to reduce uranium exposure, protect from toxic kidney effects of uranium, and reduce the risk of cancer. However, most existing private wells in the study area, generally for single family use, have not been sampled for uranium. This document presents available uranium concentration data from throughout a multi-county region, identifies data gaps, and suggests further study aimed at understanding the occurrence of uranium in groundwater.The study encompasses about 13,000 square miles (mi2) in the northeastern part of Washington with a 2010 population of about 563,000. Other than the City of Spokane, most of the study area is rural with small towns interspersed throughout the region. The study area also includes three Indian Reservations with small towns and scattered population. The area has a history of uranium exploration and mining, with two inactive uranium mines on the Spokane Indian Reservation and one smaller inactive mine on the outskirts of Spokane. Historical (1977–2016) uranium in groundwater concentration data were used to describe and illustrate the general occurrence and distribution of uranium in groundwater, as well as to identify data deficiencies. Uranium concentrations were detected at greater than 1 microgram per liter (μg/L) in 60 percent of the 2,382 historical samples (from wells and springs). Uranium concentrations ranged from less than 1 to 88,600 μg/L, and the median concentration of uranium in groundwater for all sites was 1.4 μg/L.New (2017) uranium in groundwater concentration data were obtained by sampling 13 private domestic wells for uranium in areas without recent (2000s) water-quality data. Uranium was detected in all 13 wells sampled for this study; concentrations ranged from 1.03 to 1,180 μg/L with a median of 22 μg/L. Uranium concentrations of groundwater samples from 6 of the 13 wells exceeded the MCL for uranium. Uranium concentrations in water samples from two wells were 1,130 and 1,180 μg/L, respectively; nearly 40 times the MCL.Additional data collection and analysis are needed in rural areas where self-supplied groundwater withdrawals are the primary source of water for human consumption. Of the roughly 43,000 existing water wells in the study area, only 1,755 wells, as summarized in this document, have available uranium concentration data, and some of those data are decades old. Furthermore, analysis of area groundwater quality would benefit from a more extensive chemical-analysis suite including general chemistry in order to better understand local geochemical conditions that largely govern the mobility of uranium. Although the focus of the present study is uranium, it also is important to recognize that there are other radionuclides of concern that may be present in area groundwater.

DYNAMIC PROPERTIES OF SHOCK LOADED THIN URANIUM FOILS

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

Robbins, D. L.; Kelly, A. M.; Alexander, D. J.

A series of spall experiments has been completed with thin depleted uranium targets, nominally 0.1 mm thick. The first set of uranium spall targets was cut and ground to final thickness from electro-refined, high-purity, cast uranium. The second set was rolled to final thickness from low purity uranium. The impactors for these experiments were laser-launched 0.05-mm thick copper flyers, 3 mm in diameter. Laser energies were varied to yield a range of flyer impact velocities. This resulted in varying degrees of damage to the uranium spall targets, from deformation to complete spall or separation at the higher velocities. Dynamic measurementsmore » of the uranium target free surface velocities were obtained with dual velocity interferometers. Uranium targets were recovered and sectioned after testing. Free surface velocity profiles were similar for the two types of uranium, but spall strengths (estimated from the magnitude of the pull-back signal) are higher for the high-purity cast uranium. Velocity profiles and microstructural evidence of spall from the sectioned uranium targets are presented.« less

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.

New Technique for Speciation of Uranium in Sediments Following Acetate-Stimulated Bioremediation

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

Not Available

2011-06-22

Acetate-stimulated bioremediation is a promising new technique for sequestering toxic uranium contamination from groundwater. The speciation of uranium in sediments after such bioremediation attempts remains unknown as a result of low uranium concentration, and is important to analyzing the stability of sequestered uranium. A new technique was developed for investigating the oxidation state and local molecular structure of uranium from field site sediments using X-Ray Absorption Spectroscopy (XAS), and was implemented at the site of a former uranium mill in Rifle, CO. Glass columns filled with bioactive Rifle sediments were deployed in wells in the contaminated Rifle aquifer and amendedmore » with a hexavalent uranium (U(VI)) stock solution to increase uranium concentration while maintaining field conditions. This sediment was harvested and XAS was utilized to analyze the oxidation state and local molecular structure of the uranium in sediment samples. Extended X-Ray Absorption Fine Structure (EXAFS) data was collected and compared to known uranium spectra to determine the local molecular structure of the uranium in the sediment. Fitting was used to determine that the field site sediments did not contain uraninite (UO{sub 2}), indicating that models based on bioreduction using pure bacterial cultures are not accurate for bioremediation in the field. Stability tests on the monomeric tetravalent uranium (U(IV)) produced by bioremediation are needed in order to assess the efficacy of acetate-stimulation bioremediation.« less

Health effects of uranium: new research findings.

PubMed

Brugge, Doug; Buchner, Virginia

2011-01-01

Recent plans for a nuclear renaissance in both established and emerging economies have prompted increased interest in uranium mining. With the potential for more uranium mining worldwide and a growth in the literature on the toxicology and epidemiology of uranium and uranium mining, we found it timely to review the current state of knowledge. Here, we present a review of the health effects of uranium mining, with an emphasis on newer findings (2005-2011). Uranium mining can contaminate air, water, and soil. The chemical toxicity of the metal constitutes the primary environmental health hazard, with the radioactivity of uranium a secondary concern. The update of the toxicologic evidence on uranium adds to the established findings regarding nephrotoxicity, genotoxicity, and developmental defects. Additional novel toxicologic findings, including some at the molecular level, are now emerging that raise the biological plausibility of adverse effects on the brain, on reproduction, including estrogenic effects, on gene expression, and on uranium metabolism. Historically, most epidemiology on uranium mining has focused on mine workers and radon exposure. Although that situation is still overwhelmingly true, a smaller emerging literature has begun to form around environmental exposure in residential areas near uranium mining and processing facilities. We present and critique such studies. Clearly, more epidemiologic research is needed to contribute to causal inference. As much damage is irreversible, and possibly cumulative, present efforts must be vigorous to limit environmental uranium contamination and exposure.

Uranium-bearing lignite in southwestern North Dakota

USGS Publications Warehouse

Moore, George W.; Melin, Robert E.; Kepferle, Roy C.

1954-01-01

Uranium-bearing lignite was mapped and sampled in the Bullion Butte, Sentinel Butte, HT Butte, and Chalky Buttes areas in southwestern North Dakota. The uraniferous lignite occurs at several stratigraphic positions in the Sentinel Butte member of the Fort Union formation of Paleocene age. A total of 261 samples were collected for uranium analysis from 85 localities, Lignite contained as much as 0.045 percent uranium, 10.0 percent ash, and 0.45 percent uranium in the ash was found although the average is lower. Inferred reserves for the four areas examined are estimated to be about 27 million tons of lignite in beds about 2 feet thick and containing more than 3000 tons of uranium. The lignite in beds about 2 feet thick and containing more than 3000 tons of uranium. The lignite averages more than 30 percent ash in the surface samples. The principal factor that seems to influence the uranium content of lignite beds is their stratigraphic position below the overlying rocks of the White River group of Oligocene age. All of the uranium-bearing beds closely underlie the base of the White River group. Although this relationship seems to be the controlling factor, the relative concentration of uranium may be modified by other conditions. Beds enclosed in permeable rocks are more uraniferous than beds in impermeable rocks, and thin beds have higher content of uranium than thick beds. In addition, thick lignite beds commonly have a top=preferential distribution of uranium. These and other factors suggest that the uranium is secondary and this it was introduced by ground water which had leached uranium from volcanic ash in the overlying rocks of the White River group. It is thought that the uranium is held in the lignite as part of a metallo-organic compound.

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

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

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

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

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

DOE PAGES

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

2016-06-16

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

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

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

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

Carbon diffusion in molten uranium: an ab initio molecular dynamics study

NASA Astrophysics Data System (ADS)

Garrett, Kerry E.; Abrecht, David G.; Kessler, Sean H.; Henson, Neil J.; Devanathan, Ram; Schwantes, Jon M.; Reilly, Dallas D.

2018-04-01

In this work we used ab initio molecular dynamics within the framework of density functional theory and the projector-augmented wave method to study carbon diffusion in liquid uranium at temperatures above 1600 K. The electronic interactions of carbon and uranium were described using the local density approximation (LDA). The self-diffusion of uranium based on this approach is compared with literature computational and experimental results for liquid uranium. The temperature dependence of carbon and uranium diffusion in the melt was evaluated by fitting the resulting diffusion coefficients to an Arrhenius relationship. We found that the LDA calculated activation energy for carbon was nearly twice that of uranium: 0.55 ± 0.03 eV for carbon compared to 0.32 ± 0.04 eV for uranium. Structural analysis of the liquid uranium-carbon system is also discussed.

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...

The Toxicity of Depleted Uranium

PubMed Central

Briner, Wayne

2010-01-01

Depleted uranium (DU) is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a clear and defined set of symptoms. Chronic low-dose, or subacute, exposure to depleted uranium alters the appearance of milestones in developing organisms. Adult animals that were exposed to depleted uranium during development display persistent alterations in behavior, even after cessation of depleted uranium exposure. Adult animals exposed to depleted uranium demonstrate altered behaviors and a variety of alterations to brain chemistry. Despite its reduced level of radioactivity evidence continues to accumulate that depleted uranium, if ingested, may pose a radiologic hazard. The current state of knowledge concerning DU is discussed. PMID:20195447

Geological and geochemical aspects of uranium deposits. A selected, annotated bibliography

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

Garland, P.A.; Thomas, J.M.; Brock, M.L.

1980-06-01

A bibliography of 479 references encompassing the fields of uranium and thorium geochemistry and mineralogy, geology of uranium deposits, uranium mining, and uranium exploration techniques has been compiled by the Ecological Sciences Information Center of Oak Ridge National Laboratory. The bibliography was produced for the National Uranium Resource Evaluation Program, which is funded by the Grand Junction Office of the Department of Energy. The references contained in the bibliography have been divided into the following eight subject categories: (1) geology of deposits, (2) geochemistry, (3) genesis O deposits, (4) exploration, (5) mineralogy, (6) uranium industry, (7) reserves and resources, andmore » (8) geology of potential uranium-bearing areas. All categories specifically refer to uranium and thorium; the last category contains basic geologic information concerning areas which the Grand Junction Office feels are particularly favorable for uranium deposition. The references are indexed by author, geographic location, quadrangle name, geoformational feature, taxonomic name, and keyword.« less

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.

Maps showing the distribution of uranium-deposit clusters in the Colorado Plateau uranium province

USGS Publications Warehouse

Finch, Warren I.

1991-01-01

The Colorado Palteau Uranium Province (CPUP) is defined by the distribution of uranium deposits, chiefly the sandstone-type, in upper Paleozoic and Mesozoic sedimentary rocks within the Colorado Plateau physiographic province (Granger and others, 1986).  The uranium province is bordered by widely distributed and mostly minor uranium deposits in Precambrian and Tertiary rocks and by outcrops of Tertiary extrusive and intrusive igneous rocks.  

Uranium in bone: metabolic and autoradiographic studies in the rat.

PubMed

Priest, N D; Howells, G R; Green, D; Haines, J W

1982-03-01

The distribution and retention of intravenously injected hexavalent uranium-233 in the skeleton of the female rat has been investigated using a variety of autoradiographic and radiochemical techniques. These showed that approximately one third of the injected uranium is deposited in the skeleton where it is retained with an initial biological half-time of approximately 40 days. The studies also showed that: 1 Uranium is initially deposited onto all types of bone surface, but preferentially onto those that are accreting. 2 Uranium is deposited in the calcifying zones of skeletal cartilage. 3 Bone accretion results in the burial of surface deposits of uranium. 4 Bone resorption causes the removal of uranium from surfaces. 5 Resorbed uranium is not retained by osteoclasts and macrophages in the bone marrow. 6 Uranium removed from bone surfaces enters the bloodstream where most is either redeposited in bone or excreted via the kidneys. 7 The recycling of resorbed uranium within the skeleton tends to produce a uniform level of uranium contamination throughout mineralized bone. These results are taken to indicate that uranium deposition in bone shares characteristics in common with both the 'volume-seeking radionuclides' typified by the alkaline earth elements and with the 'bone surface-seeking radionuclides' typified by plutonium.

Rolling Process Modeling Report: Finite-Element Prediction of Roll Separating Force and Rolling Defects

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

Soulami, Ayoub; Lavender, Curt A.; Paxton, Dean M.

2014-04-23

Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum (U-10Mo) alloy plate-type fuel for the U.S. high-performance research reactors. This work supports the Convert Program of the U.S. Department of Energy’s National Nuclear Security Administration (DOE/NNSA) Global Threat Reduction Initiative. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll separating forces and rolling defects. Simulations were performed using a finite-element model developed using the commercial code LS-Dyna. Simulations of the hot rolling of U-10Mo coupons encapsulated in low-carbon steel have been conducted following two different schedules. Model predictions ofmore » the roll-separation force and roll-pack thicknesses at different stages of the rolling process were compared with experimental measurements. This report discusses various attributes of the rolled coupons revealed by the model (e.g., dog-boning and thickness non-uniformity).« less

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

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

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« 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.

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

Uranium hydrogeochemical and stream sediment reconnaissance of the Albuquerque NTMS Quadrangle, New Mexico, including concentrations of forty-three additional elements

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

Maassen, L.W.; Bolivar, S.L.

1979-06-01

The Los Alamos Scientific Laboratory conducted a hydrogeochemical and stream sediment reconnaissance for uranium. Totals of 408 water and 1538 sediment samples were collected from 1802 locations over a 20 100-km/sup 2/ area at an average density of one location per 11 km/sup 2/. Water samples were collected from springs, wells, and streams; sediments samples were collected predominantly from streams, but also from springs. All water samples were analyzed for uranium and 12 other elements. Sediment samples were analyzed for uranium and 42 additional elements. The uranium concentrations in water samples range from below the detection limit of 0.02 ppBmore » to 194.06 ppB. The mean uranium concentration for all water types containing < 40 ppB uranium is 1.98 ppB. Six samples contained uranium concentrations > 40.00 ppB. Well waters have the highest mean uranium concentration; spring waters have the lowest. Clusters of water samples that contain anomalous uranium concentrations are delineated in nine areas. Sediments collected from the quadrangle have uranium concentrations that range between 0.63 ppM and 28.52 ppM, with a mean for all sediments of 3.53 ppM. Eight areas containing clusters of sediments with anomalous uranium concentrations are delineated. One cluster contains sample locations within the Ambrosia Lake uranium district. Five clusters of sediment samples with anomalous uranium concentrations were collected from streams that drain the Jemez volcanic field. Another cluster defines an area just northeast of Albuquerque where streams drain Precambrian rocks, predominantly granites, of the Sandia Mountains. The last cluster, consisting of spring sediments from Mesa Portales, was collected near the contact of the Tertiary Ojo Alamo sandstone with underlying Cretaceous sediments. Sediments from these springs exhibit some of the highest uranium values reported and are associated with high uranium/thorium ratios.« less

31 CFR 540.318 - Uranium Hexafluoride (UF6).

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Uranium Hexafluoride (UF6). 540.318... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.318 Uranium Hexafluoride (UF6). The term uranium...

49 CFR 173.426 - Excepted packages for articles containing natural uranium or thorium.

Code of Federal Regulations, 2013 CFR

2013-10-01

... uranium or thorium. 173.426 Section 173.426 Transportation Other Regulations Relating to Transportation....426 Excepted packages for articles containing natural uranium or thorium. A manufactured article in which the sole Class 7 (radioactive) material content is natural uranium, unirradiated depleted uranium...

31 CFR 540.318 - Uranium Hexafluoride (UF6).

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Uranium Hexafluoride (UF6). 540.318... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.318 Uranium Hexafluoride (UF6). The term uranium...

31 CFR 540.318 - Uranium Hexafluoride (UF6).

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Uranium Hexafluoride (UF6). 540.318... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.318 Uranium Hexafluoride (UF6). The term uranium...

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 540.318 - Uranium Hexafluoride (UF6).

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Uranium Hexafluoride (UF6). 540.318... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.318 Uranium Hexafluoride (UF6). The term uranium...

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...

49 CFR 173.426 - Excepted packages for articles containing natural uranium or thorium.

Code of Federal Regulations, 2014 CFR

2014-10-01

... uranium or thorium. 173.426 Section 173.426 Transportation Other Regulations Relating to Transportation....426 Excepted packages for articles containing natural uranium or thorium. A manufactured article in which the sole Class 7 (radioactive) material content is natural uranium, unirradiated depleted uranium...

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...

49 CFR 173.426 - Excepted packages for articles containing natural uranium or thorium.

Code of Federal Regulations, 2012 CFR

2012-10-01

... uranium or thorium. 173.426 Section 173.426 Transportation Other Regulations Relating to Transportation....426 Excepted packages for articles containing natural uranium or thorium. A manufactured article in which the sole Class 7 (radioactive) material content is natural uranium, unirradiated depleted uranium...

31 CFR 540.318 - Uranium Hexafluoride (UF6).

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium Hexafluoride (UF6). 540.318... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.318 Uranium Hexafluoride (UF6). The term uranium...

31 CFR 540.309 - Natural uranium.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Natural uranium. 540.309 Section 540.309 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... REGULATIONS General Definitions § 540.309 Natural uranium. The term natural uranium means uranium found in...

31 CFR 540.309 - Natural uranium.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Natural uranium. 540.309 Section 540.309 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... REGULATIONS General Definitions § 540.309 Natural uranium. The term natural uranium means uranium found in...

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)

Separation of uranium from (Th,U)O.sub.2 solid solutions

DOEpatents

Chiotti, Premo; Jha, Mahesh Chandra

1976-09-28

Uranium is separated from mixed oxides of thorium and uranium by a pyrometallurgical process in which the oxides are mixed with a molten chloride salt containing thorium tetrachloride and thorium metal which reduces the uranium oxide to uranium metal which can then be recovered from the molten salt. The process is particularly useful for the recovery of uranium from generally insoluble high-density sol-gel thoria-urania nuclear reactor fuel pellets.

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.

URANIUM RECOVERY PROCESS

DOEpatents

Bailes, R.H.; Long, R.S.; Olson, R.S.; Kerlinger, H.O.

1959-02-10

A method is described for recovering uranium values from uranium bearing phosphate solutions such as are encountered in the manufacture of phosphate fertilizers. The solution is first treated with a reducing agent to obtain all the uranium in the tetravalent state. Following this reduction, the solution is treated to co-precipitate the rcduced uranium as a fluoride, together with other insoluble fluorides, thereby accomplishing a substantially complete recovery of even trace amounts of uranium from the phosphate solution. This precipitate usually takes the form of a complex fluoride precipitate, and after appropriate pre-treatment, the uranium fluorides are leached from this precipitate and rccovered from the leach solution.

Carbon diffusion in molten uranium: an ab initio molecular dynamics study

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

Garrett, Kerry E.; Abrecht, David G.; Kessler, Sean H.

In this work we used ab initio molecular dynamics (AIMD) within the framework of density functional theory (DFT) and the projector-augmented wave (PAW) method to study carbon diffusion in liquid uranium at temperatures above 1600 K. The electronic interactions of carbon and uranium were described using the local density approximation (LDA). The self-diffusion of uranium based on this approach is compared with literature computational and experimental results for liquid uranium. The temperature dependence of carbon and uranium diffusion in the melt was evaluated by fitting the resulting diffusion coefficients to an Arrhenius relationship. We found that the LDA calculated activationmore » energy for carbon was nearly twice that of uranium: 0.55±0.03 eV for carbon compared to 0.32±0.04 eV for uranium. Structural analysis of the liquid uranium-carbon system is also discussed.« less

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.

Fermentation and Hydrogen Metabolism Affect Uranium Reduction by Clostridia

DOE PAGES

Gao, Weimin; Francis, Arokiasamy J.

2013-01-01

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

Incorporation of Uranium: II. Distribution of Uranium Absorbed through the Lungs and the Skin

PubMed Central

Walinder, G.; Fries, B.; Billaudelle, U.

1967-01-01

In experiments on mice, rabbits, and piglets the distribution of uranium was studied at different times after exposure. Uranium was administered by inhalation (mice) and through the skin (rabbits and piglets). These investigations show that the uptakes of uranium in different organs of the three species are highly dependent on the amounts administered. There seems to be a saturation effect in the spleen and bone tissue whenever the uranium concentration in the blood exceeds a certain level. The effect in the kidney is completely different. If, in a series of animals, the quantity of uranium is continuously increased, the uptakes by the kidneys increase more rapidly than the quantities administered. This observation seems to be consistent with the toxic effects of uranium on the capillary system in the renal cortex. Polyphloretin phosphate, a compound which reduces permeability, was investigated with respect to its effect on the uptake of uranium deposited in skin wounds in rabbits and piglets. It significantly reduced the absorption of uranium, even from depots in deep wounds. The findings are discussed with reference to the routine screening of persons exposed to uranium at AB Atomenergi. Images PMID:6073090

Rapid Dissolution of Soluble Uranyl Phases in Arid, Mine-Impacted Catchments Near Church Rock, NM

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

deLemos, J.L.; Bostick, B.C.; Quicksall, A.N.

2009-05-14

We tested the hypothesis that runoff of uranium-bearing particles from mining waste disposal areas was a significant mechanism for redistribution of uranium in the northeastern part of the Upper Puerco River watershed (New Mexico). However, our results were not consistent with this hypothesis. Analysis of >100 sediment and suspended sediment samples collected adjacent to and downstream from uranium source areas indicated that uranium levels in the majority of the samples were not elevated above background. Samples collected within 50 m of a known waste disposal site were subjected to detailed geochemical characterization. Uranium in these samples was found to bemore » highly soluble; treatment with synthetic pore water for 24 h caused dissolution of 10-50% of total uranium in the samples. Equilibrium uranium concentrations in pore water were >4.0 mg/L and were sustained in repeated wetting events, effectively depleting soluble uranium from the solid phase. The dissolution rate of uranium appeared to be controlled by solid-phase diffusion of uranium from within uranium-bearing mineral particles. X-ray adsorption spectroscopy indicated the presence of a soluble uranyl silicate, and possibly a uranyl phosphate. These phases were exhausted in transported sediment suggesting that uranium was readily mobilized from sediments in the Upper Puerco watershed and transported in the dissolved load. These results could have significance for uranium risk assessment as well as mining waste management and cleanup efforts.« less

Comparison of the Kinetic Rate Law Parameters for the Dissolution of Natural and Synthetic Autunite in the Presence of Aqueous Bicarbonate Ions

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

Gudavalli, Ravi; Katsenovich, Yelena; Wellman, Dawn M.

2013-08-02

Bicarbonate is one of the most significant components within the uranium geochemical cycle. In aqueous solutions, bicarbonate forms strong complexes with uranium. As such, aqueous bicarbonate may significantly increase the rate of uranium release from uranium minerals. Quantifying the relationship of aqueous bicarbonate concentration to the rate of uranium release during dissolution is critical to understanding the long-term fate of uranium within the environment. Single-pass flow-through (SPTF) experiments were conducted to estimate the rate of uranium release from Na meta-autunite as a function of bicarbonate (0.0005-0.003 M) under the pH range of 6-11 and a temperature range of 5-60oC. Consistentmore » with the results of previous investigation, the rate of uranium release exhibited minimal dependency on temperature; but were strongly dependent on pH. Increasing aqueous bicarbonate concentrations afforded comparable increases in the rate of release of uranium. Most notably under low pH conditions the aqueous bicarbonate resulted in up to 370 fold increases in the rate of uranium release in relative to the rate of uranium release in the absence of bicarbonate. However, the effect of aqueous bicarbonate on the release of uranium was significantly less under higher pH conditions. It is postulated that at high pH values, surface sites are saturated with carbonate, thus the addition of more bicarbonate would have less effect on uranium release.« less

Rapid Dissolution of Soluble Uranyl Phases in Arid, Mine-Impacted Catchments near Church Rock, NM

PubMed Central

DELEMOS, JAMIE L.; BOSTICK, BENJAMIN C.; QUICKSALL, ANDREW N.; LANDIS, JOSHUA D.; GEORGE, CHRISTINE C.; SLAGOWSKI, NAOMI L.; ROCK, TOMMY; BRUGGE, DOUG; LEWIS, JOHNNYE; DURANT, JOHN L.

2008-01-01

We tested the hypothesis that runoff of uranium-bearing particles from mining waste disposal areas was a significant mechanism for redistribution of uranium in the northeastern part of the Upper Puerco River watershed (New Mexico). However, our results were not consistent with this hypothesis. Analysis of >100 sediment and suspended sediment samples collected adjacent to and downstream from uranium source areas indicated that uranium levels in the majority of the samples were not elevated above background. Samples collected within 50 m of a known waste disposal site were subjected to detailed geochemical characterization. Uranium in these samples was found to be highly soluble; treatment with synthetic pore water for 24 h caused dissolution of 10–50% of total uranium in the samples. Equilibrium uranium concentrations in pore water were >4.0 mg/L and were sustained in repeated wetting events, effectively depleting soluble uranium from the solid phase. The dissolution rate of uranium appeared to be controlled by solid-phase diffusion of uranium from within uranium-bearing mineral particles. X-ray adsorption spectroscopy indicated the presence of a soluble uranyl silicate, and possibly a uranyl phosphate. These phases were exhausted in transported sediment suggesting that uranium was readily mobilized from sediments in the Upper Puerco watershed and transported in the dissolved load. These results could have significance for uranium risk assessment as well as mining waste management and cleanup efforts. PMID:18589950

Biosorption of uranium by Pseudomonas aeruginosa strain CSU: Characterization and comparison studies

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

Hu, M.Z.C.; Norman, J.M.; Faison, B.D.

1996-07-20

Pseudomonas aeruginosa strain CSU, a nongenetically engineered bacterial strain known to bind dissolved hexavalent uranium (as UO{sub 2}{sup 2+} and/or its cationic hydroxo complexes) was characterized with respect to its sorptive activity. The uranium biosorption equilibrium could be described by the Langmuir isotherm. The rate of uranium adsorption increased following permeabilization of the outer and/or cytoplasmic membrane by organic solvents such as acetone. P. aeruginosa CSU biomass was significantly more sorptive toward uranium than certain novel, patented biosorbents derived from algal or fungal biomass sources. P. aeruginosa CSU biomass was also competitive with commercial cation-exchange resins, particularly in the presencemore » of dissolved transition metals. Uranium binding by P. aeruginosa CSU was clearly pH dependent. Uranium loading capacity increased with increasing pH under acidic conditions, presumably as a function of uranium speciation and due to the H{sup +} competition at some binding sites. Nevertheless, preliminary evidence suggests that this microorganism is also capable of binding anionic hexavalent uranium complexes. Ferric iron was a strong inhibitor of uranium binding to P. aeruginosa CSU biomass, and the presence of uranium also decreased the Fe{sup 3+} loading when the biomass was not saturated with Fe{sup 3+}. Thus, a two-state process in which iron and uranium are removed in consecutive steps was proposed for efficient use of the biomass as a biosorbent in uranium removal from mine wastewater, especially acidic leachates.« less

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.

Challenges dealing with depleted uranium in Germany - Reuse or disposal

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

Moeller, Kai D.

2007-07-01

During enrichment large amounts of depleted Uranium are produced. In Germany every year 2.800 tons of depleted uranium are generated. In Germany depleted uranium is not classified as radioactive waste but a resource for further enrichment. Therefore since 1996 depleted Uranium is sent to ROSATOM in Russia. However it still has to be dealt with the second generation of depleted Uranium. To evaluate the alternative actions in case a solution has to be found in Germany, several studies have been initiated by the Federal Ministry of the Environment. The work that has been carried out evaluated various possibilities to dealmore » with depleted uranium. The international studies on this field and the situation in Germany have been analyzed. In case no further enrichment is planned the depleted uranium has to be stored. In the enrichment process UF{sub 6} is generated. It is an international consensus that for storage it should be converted to U{sub 3}O{sub 8}. The necessary technique is well established. If the depleted Uranium would have to be characterized as radioactive waste, a final disposal would become necessary. For the planned Konrad repository - a repository for non heat generating radioactive waste - the amount of Uranium is limited by the licensing authority. The existing license would not allow the final disposal of large amounts of depleted Uranium in the Konrad repository. The potential effect on the safety case has not been roughly analyzed. As a result it may be necessary to think about alternatives. Several possibilities for the use of depleted uranium in the industry have been identified. Studies indicate that the properties of Uranium would make it useful in some industrial fields. Nevertheless many practical and legal questions are open. One further option may be the use as shielding e.g. in casks for transport or disposal. Possible techniques for using depleted Uranium as shielding are the use of the metallic Uranium as well as the inclusion in concrete. Another possibility could be the use of depleted uranium for the blending of High enriched Uranium (HEU) or with Plutonium to MOX-elements. (authors)« less

Uranium mineralization and unconformities: how do they correlate? - A look beyond the classic unconformity-type deposit model?

NASA Astrophysics Data System (ADS)

Markwitz, Vanessa; Porwal, Alok; Campbell McCuaig, T.; Kreuzer, Oliver P.

2010-05-01

Uranium deposits are usually classified based on the characteristics of their host rocks and geological environments (Dahlkamp, 1993; OECD/NEA Red Book and IAEA, 2000; Cuney, 2009). The traditional unconformity-related deposit types are the most economical deposits in the world, with the highest grades amongst all uranium deposit types. In order to predict undiscovered uranium deposits, there is a need to understand the spatial association of uranium mineralization with structures and unconformities. Hydrothermal uranium deposits develop by uranium enriched fluids from source rocks, transported along permeable pathways to their depositional environment. Unconformities are not only separating competent from incompetent sequences, but provide the physico-chemical gradient in the depositional environment. They acted as important fluid flow pathways for uranium to migrate not only for surface-derived oxygenated fluids, but also for high oxidized metamorphic and magmatic fluids, dominated by their geological environment in which the unconformities occur. We have carried out comprehensive empirical spatial analyses of various types of uranium deposits in Australia, and first results indicate that there is a strong spatial correlation between unconformities and uranium deposits, not only for traditional unconformity-related deposits but also for other styles. As a start we analysed uranium deposits in Queensland and in particular Proterozoic metasomatic-related deposits in the Mount Isa Inlier and Late Carboniferous to Early Permian volcanic-hosted uranium occurrences in Georgetown and Charters Towers Regions show strong spatial associations with contemporary and older unconformities. The Georgetown Inlier in northern Queensland consists of a diverse range of rocks, including Proterozoic and early Palaeozoic metamorphic rocks and granites and late Palaeozoic volcanic rocks and related granites. Uranium-molybdenum (+/- fluorine) mineralization in the Georgetown inlier varies from strata- to structure-bound and occurs above regional unconformities. The Proterozoic basins in the Mount Isa Inlier rest unconformably on Palaeoproterozoic basement accompanied by volcanic and igneous rocks, which were deformed and metamorphosed in the Mesoproterozoic. Uranium occurrences in the Western Succession of Mount Isa are either hosted in clastic metasediments or mafic volcanics that belong to the Palaeoproterozoic Eastern Creek Volcanics. Uranium and vanadium mineralization occur in metasomatised and hematite-magnetite-carbonate alteration zones, bounded by major faults and regional unconformities. The results of this study highlight the importance of unconformities in uranium minerals systems as possible fluid pathways and/or surfaces of physico-chemical contrast that could have facilitated the precipitation of uranium, not only in classical unconformity style uranium deposits but in several other styles of uranium mineralization as well. References Cuney, M., 2009. The extreme diversity of uranium deposits. Mineralium Deposita, 44, 3-9. Dahlkamp, F. J., 1993. Uranium ore deposits. Springer, Berlin, p 460. OECD / NEA Red Book & IAEA, 2000. Uranium 1999: Resources, Production and Demand. OECD Nuclear Energy Agency and International Atomic Energy Agency, Paris.

Natural uranium impairs the differentiation and the resorbing function of osteoclasts.

PubMed

Gritsaenko, Tatiana; Pierrefite-Carle, Valérie; Lorivel, Thomas; Breuil, Véronique; Carle, Georges F; Santucci-Darmanin, Sabine

2017-04-01

Uranium is a naturally occurring radionuclide ubiquitously present in the environment. The skeleton is the main site of uranium long-term accumulation. While it has been shown that natural uranium is able to perturb bone metabolism through its chemical toxicity, its impact on bone resorption by osteoclasts has been poorly explored. Here, we examined for the first time in vitro effects of natural uranium on osteoclasts. The effects of uranium on the RAW 264.7 monocyte/macrophage mouse cell line and primary murine osteoclastic cells were characterized by biochemical, molecular and functional analyses. We observed a cytotoxicity effect of uranium on osteoclast precursors. Uranium concentrations in the μM range are able to inhibit osteoclast formation, mature osteoclast survival and mineral resorption but don't affect the expression of the osteoclast gene markers Nfatc1, Dc-stamp, Ctsk, Acp5, Atp6v0a3 or Atp6v0d2 in RAW 274.7 cells. Instead, we observed that uranium induces a dose-dependent accumulation of SQSTM1/p62 during osteoclastogenesis. We show here that uranium impairs osteoclast formation and function in vitro. The decrease in available precursor cells, as well as the reduced viability of mature osteoclasts appears to account for these effects of uranium. The SQSTM1/p62 level increase observed in response to uranium exposure is of particular interest since this protein is a known regulator of osteoclast formation. A tempting hypothesis discussed herein is that SQSTM1/p62 dysregulation contributes to uranium effects on osteoclastogenesis. We describe cellular and molecular effects of uranium that potentially affect bone homeostasis. Copyright © 2017 Elsevier B.V. All rights reserved.

Bone as a Possible Target of Chemical Toxicity of Natural Uranium in Drinking Water

PubMed Central

Kurttio, Päivi; Komulainen, Hannu; Leino, Aila; Salonen, Laina; Auvinen, Anssi; Saha, Heikki

2005-01-01

Uranium accumulates in bone, affects bone metabolism in laboratory animals, and when ingested in drinking water increases urinary excretion of calcium and phosphate, important components in the bone structure. However, little is known about bone effects of ingested natural uranium in humans. We studied 146 men and 142 women 26–83 years of age who for an average of 13 years had used drinking water originating from wells drilled in bedrock, in areas with naturally high uranium content. Biochemical indicators of bone formation were serum osteocalcin and amino-terminal propeptide of type I procollagen, and a marker for bone resorption was serum type I collagen carboxy-terminal telopeptide (CTx). The primary measure of uranium exposure was uranium concentration in drinking water, with additional information on uranium intake and uranium concentration in urine. The data were analyzed separately for men and women with robust regression (which suppresses contributions of potential influential observations) models with adjustment for age, smoking, and estrogen use. The median uranium concentration in drinking water was 27 μg/L (interquartile range, 6–116 μg/L). The median of daily uranium intake was 36 μg (7–207 μg) and of cumulative intake 0.12 g (0.02–0.66 g). There was some suggestion that elevation of CTx (p = 0.05) as well as osteocalcin (p = 0.19) could be associated with increased uranium exposure (uranium in water and intakes) in men, but no similar relationship was found in women. Accordingly, bone may be a target of chemical toxicity of uranium in humans, and more detailed evaluation of bone effects of natural uranium is warranted. PMID:15626650

Integrated geophysical investigations for the delineation of source and subsurface structure associated with hydro-uranium anomaly: A case study from South Purulia Shear Zone (SPSZ), India

NASA Astrophysics Data System (ADS)

Sharma, S. P.; Biswas, A.

2012-12-01

South Purulia Shear Zone (SPSZ) is an important region for prospecting of uranium mineralization. Geological studies and hydro-uranium anomaly suggest the presence of Uranium deposit around Raghunathpur village which lies about 8 km north of SPSZ. However, detailed geophysical investigations have not been carried out in this region for investigation of uranium mineralization. Since surface signature of uranium mineralization is not depicted near the location, a deeper subsurface source is expected for hydro uranium anomaly. To delineate the subsurface structure and to investigate the origin of hydro-uranium anomaly present in the area, Vertical Electrical Sounding (VES) using Schlumberger array and Gradient Resistivity Profiling (GRP) were performed at different locations along a profile perpendicular to the South Purulia Shear Zone. Apparent resistivity computed from the measured sounding data at various locations shows a continuously increasing trend. As a result, conventional apparent resistivity data is not able to detect the possible source of hydro uranium anomaly. An innovative approach is applied which depicts the apparent conductivity in the subsurface revealed a possible connection from SPSZ to Raghunathpur. On the other hand resistivity profiling data suggests a low resistive zone which is also characterized by low Self-Potential (SP) anomaly zone. Since SPSZ is characterized by the source of uranium mineralization; hydro-uranium anomaly at Raghunathpur is connected with the SPSZ. The conducting zone has been delineated from SPSZ to Raghunathpur at deeper depths which could be uranium bearing. Since the location is also characterized by a low gravity and high magnetic anomaly zone, this conducting zone is likely to be mineralized zone. Keywords: Apparent resistivity; apparent conductivity; Self Potential; Uranium mineralization; shear zone; hydro-uranium anomaly.

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)

DECONTAMINATION OF URANIUM

DOEpatents

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

1958-02-01

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

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

PubMed

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

2013-07-02

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

Hydrogeochemical and stream sediment reconnaissance basic data for Brownsville-McAllen NTMS Quadrangles, Texas

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

Not Available

1980-09-30

Results of a reconnaissance geochemical survey of the Brownsville-McAllen Quadrangles, Texas are reported. Field and laboratory data are presented for 427 groundwater and 171 stream sediment samples. Statistical and areal distributions of uranium and possible uranium-related variables are displayed. Pertinent geologic factors which may be of significance in evaluating the potential for uranium mineralization are briefly discussed. Groundwater data indicate the most promising area for potential uranium mineralization occurs in the northwestern section of the quadrangles (Jim Hogg, Starr, and Zapata Counties), where waters are derived from the Catahoula Formation. These groundwaters have high concentrations of uranium, uranium associated elements,more » and low values for specific conductance. Another area with high uranium concentrations is in the southeastern portion of the survey area (Hidalgo, Cameron, and Willacy Counties). Shallow wells <10 m (30 ft) are numerous in this area and high specific conductance values may indicate contamination from extensive fertilization. Stream sediment data for the survey does not indicate an area favorable for uranium mineralization. Anomalous acid soluble uranium values in the southeastern area (Hidalgo, Cameron, and Willacy Counties) can be attributed to phosphate fertilizer contamination. Four samples in the western part of the area (western Starr County) have anomalously high total uranium values and low acid soluble uranium values, indicating the uranium may be contained in resistate minerals.« less

Uranium hydrogeochemical and stream sediment reconnaissance of the Newcastle NTMS quadrangle, Wyoming, including concentrations of forty-two additional elements

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

Goff, S.J.; Sandoval, W.F.; Gallimore, D.L.

1980-06-01

During the summer and fall of 1977, 533 water and 1226 sediment samples were collected from 1740 locations within the 18,000 km/sup 2/ area of the Newcastle quadrangle, Wyoming. Water samples were collected from wells and springs; sediment samples were collected from stream channels and from springs. Each water sample was analyzed for uranium, and each sediment sample was analyzed for 43 elements, including uranium and thorium. Uranium concentrations in water samples range from below the detection limit of 0.02 ppB to 702.26 ppB and have a median of 1.73 ppB and a mean of 11.76 ppB. Water samples containingmore » high uranium concentrations (>20 ppB) generally are associated with known uranium mining activity or units known to be uranium bearing. About one-third of the water samples containing high uranium concentrations were collected from locations within the Pumpkin Buttes and Turnercrest-Ross Districts. Nearly half of the water samples containing high uranium concentrations were collected from locations just west of the Monument Hill and Highland Flats-Box Creek Districts. Similar anomalous uranium concentrations in this region have been reported updip from Exxon's Highland uranium deposits. High uranium concentrations were also found associated with the Lance Creek-Old Woman Anticline District.« 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.

31 CFR 540.315 - Uranium-235 (U235).

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

31 CFR 540.315 - Uranium-235 (U235).

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

49 CFR 173.434 - Activity-mass relationships for uranium and natural thorium.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 2 2014-10-01 2014-10-01 false Activity-mass relationships for uranium and....434 Activity-mass relationships for uranium and natural thorium. The table of activity-mass relationships for uranium and natural thorium are as follows: Thorium and uranium enrichment 1(Wt% 235 U present...

49 CFR 173.434 - Activity-mass relationships for uranium and natural thorium.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 2 2012-10-01 2012-10-01 false Activity-mass relationships for uranium and....434 Activity-mass relationships for uranium and natural thorium. The table of activity-mass relationships for uranium and natural thorium are as follows: Thorium and uranium enrichment 1(Wt% 235 U present...

49 CFR 173.434 - Activity-mass relationships for uranium and natural thorium.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 2 2013-10-01 2013-10-01 false Activity-mass relationships for uranium and....434 Activity-mass relationships for uranium and natural thorium. The table of activity-mass relationships for uranium and natural thorium are as follows: Thorium and uranium enrichment 1(Wt% 235 U present...

31 CFR 540.315 - Uranium-235 (U235).

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

31 CFR 540.315 - Uranium-235 (U235).

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

31 CFR 540.315 - Uranium-235 (U235).

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

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

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.

Ramie (Boehmeria nivea)'s uranium bioconcentration and tolerance attributes.

PubMed

Wang, Wei-Hong; Luo, Xue-Gang; Liu, Lai; Zhang, Yan; Zhao, Hao-Zhou

2018-04-01

The authors sampled and analyzed 15 species of dominant wild plants in Huanan uranium tailings pond in China, whose tailings' uranium contents were 3.21-120.52 μg/g. Among the 15 species of wild plants, ramie (Boehmeria nivea) had the strongest uranium bioconcentration and transfer capacities. In order to study the uranium bioconcentration and tolerance attributes of ramie in detail, and provide a reference for the screening remediation plants to phytoremedy on a large scale in uranium tailings pond, a ramie cultivar Xiangzhu No. 7 pot experiment was carried out. We found that both wild ramie and Xiangzhu No. 7 could bioconcentrate uranium, but there were two differences. One was wild ramie's shoots bioconcentrated uranium up to 20 μg/g (which can be regarded as the critical content value of the shoot of uranium hyperaccumulator) even the soil uranium content was as low as 5.874 μg/g while Xiangzhu No. 7's shoots could reach 20 μg/g only when the uranium treatment concentrations were 275 μg/g or more; the other was that all the transfer factors of 3 wild samples were >1, and the transfer factors of 27 out of 28 pot experiment samples were <1. Probably wild ramie was a uranium hyperaccumulator. Xiangzhu No. 7 satisfied the needs of uranium hyperaccumulator on accumulation capability, tolerance capability, bioconcentration factor, but not transfer capability, so Xiangzhu No. 7 was not a uranium hyperaccumulator. We analyzed the possible reasons why there were differences in the uranium bioconcentration and transfer attributes between wild ramie and Xiangzhu No. 7., and proposed the direction for further research. In our opinion, both the plants which bioconcentrate contaminants in the shoots and roots can act as phytoextractors. Although Xiangzhu No. 7's biomass and accumulation of uranium were concentrated on the roots, the roots were small in volume and easy to harvest. And Xiangzhu No. 7's cultivating skills and protection measures had been developed very well. Xiangzhu No. 7's whole bioconcentration factors and the roots' bioconcentration factors, which were 1.200-1.834 and 1.460-2.341, respectively, increased with the increases of uranium contents of pot soil when the soil's uranium contents are 25-175 μg/g, so it can act as a potential phytoextractor when Huanan uranium tailings pond is phytoremediated. Copyright © 2018. Published by Elsevier Ltd.

Electron correlation and relativity of the 5f electrons in the U-Zr alloy system

NASA Astrophysics Data System (ADS)

Söderlind, P.; Sadigh, B.; Lordi, V.; Landa, A.; Turchi, P. E. A.

2014-01-01

We address a recently communicated conception that spin-orbit interaction and strong electron correlations are important for the metal fuel U-Zr system. Here, we show that (i) relativistic effects only marginally correct the uranium metal equation-of-state and (ii) addition of onsite Coulomb repulsion leads to an unphysical magnetic ground state of the body-centered cubic (γ) phase and a grossly overestimated equilibrium volume. Consequently, LSDA + U is deemed unsuitable for describing the electronic structure of the U-Zr system. Recently, Xiong et al. [1] reported on thermodynamic modeling of the U-Zr system motivated by its potential as a nuclear fuel for fast breeder reactors. This work [1] came on the heels of another report by Landa et al. [2] on the same system, but with very different results for the formation enthalpies and ultimate conclusion on the U-Zr phase diagram. The authors [1] argue that their calculated energetics are significantly more accurate than that by Landa et al. [2], and they further attribute the difference to strong electron correlations and the relativistic spin-orbit interaction.In the present letter we show that uranium metal, and thus the U-Zr metal nuclear fuel system, possess weakly correlated electrons that are adequately described within density-functional theory in the generalized gradient approximation, and that addition of onsite Coulomb repulsion using the LSDA + U formalism leads to finite magnetization of the γ phase in contradiction to experiments. Furthermore, we show that spin-orbit interaction is quite weak in uranium metal and that its inclusion will not significantly change the chemical bonding and formation enthalpies.In order to illustrate our arguments, we perform comparative electronic-structure calculations using the full-potential linear augmented plane-wave (FPLAPW) method and the projector augmented plane-wave (PAW) method as implemented in the Wien2K [3] and VASP [4] codes. The Wien2K computations are set up with an APW + lo basis for the expansion of the wave functions within the muffin-tin spheres (with radius RMT = 2.5 a.u.) in partial waves with angular momenta up to l = 3, and an LAPW basis for all higher angular momenta up to l = 10. The plane-wave cutoff (Kmax) for the expansion of the wave functions in the interstitial region is chosen such that RMT × Kmax = 10. We apply the LSDA + U scheme proposed by Anisimov et al. [5] (Wien2K) and Dudarev et al. [6] (VASP) to the uranium f orbitals, which approximately corrects for their electron self interaction. An effective Ueff = U - J is chosen to be 2 eV (J = 0), which appears to be realistic for uranium systems [7]. The spin-orbit interaction is included using the second-variation method with scalar-relativistic orbitals as basis. This basis includes all Eigen states with energy less than 70 eV. For reason to improve the description of the relativistic orbitals, the p1/2 local orbitals are added to the basis set. For actinide metals, this technique for the spin-orbit coupling equals, with good approximation, that of the complete four-spinor Dirac formalism [8-10]. All calculations use a 12 × 12 × 12 Monkhorst-Pack k-point grid and a plane-wave cutoff of 23 Ry.In Table 1 we show our calculated equilibrium volumes (V) and bulk moduli (B) obtained with and without spin-orbit coupling (SOC) for bcc (γ) uranium metal using the Wien2K (VASP) codes. (The other component, Zr, is a light metal where relativistic effects are not important). The changes in V and B due to SOC are indeed quite small, consistent with results from previous studies [9,11], and within the scatter of the experimental data. The reason why the volume expands slightly is that the separation of the 5f5/2 and 5f7/2 states, due to spin-orbit coupling, weakens the cohesion of the bonding electrons. The separation is very small, as seen in Fig. 1[11] where we plot the total electronic density-of-states (DOS) for α-U with and without SOC. Most of the difference occurs well above the Fermi level, and this also explains why SOC is more important for the heavier metals Pu and Am [12]. Because of the very small influence of SOC on V and B for γ-U, one may argue that also the formation enthalpies for the U-Zr system, that substantially depend on these properties, are insensitive to the SOC, contrary to the conclusion by Xiong et al. [1].Next, we address the issue of strong electron correlation and the need for an LSDA + U type of approach for γ-uranium. Searching through the literature we could not find any publication where the LSDA + U methodology was applied to uranium metal, except that of [1], but for uranium oxides they are plentiful. However, we found that it has been considered for its nearest-neighbor metals Pa [13] and Np [14] (U lies between them in the periodic table). These publications suggest that LSDA + U are not necessary for the metals while for their oxides it is relevant, although the methodology itself is phenomenological. Focusing first on the uranium ground state, orthorhombic α-U, equilibrium and structural properties [11], elastic constants [11,15], phonon spectra [16], various defects [17-19], and even subtle electronic-structure details related to the charge-density waves [20] are all satisfactorily described within conventional DFT. These results clearly imply that LSDA + U are not a relevant or necessary scheme for α-uranium. How about the high-temperature γ phase?The γ phase is stable at temperatures above 1100 K and it has a significantly larger volume than α-U, see Table 1. One may suspect that this is due to f-electron localization (strong f-electron correlation), but it actually stems from normal thermal volume expansion.In Fig. 2 we show the calculated α-U thermal volume expansion, using a parameter-free Debye-Grüneisen quasi-harmonic theory [21,22]. Also included are experimental data points taken from Donohue [23]. The good agreement with experiment for the linear expansion coefficient (αL) at room temperature gives us confidence that the quasi-harmonic model is reasonable. The thermal volume expansion shown in the figure further demonstrates (i) that at room temperature α-U has nearly identical volume to Mo-stabilized γ-U (extrapolated to zero Mo content) at the same temperature [23]; and (ii) at about 1100 K the calculated α-U volume is consistent with that measured [23] for the γ phase. Hence, the difference between α and γ volumes is completely explained by thermal expansion with no need for additional f-electron localization. This conclusion is further underscored by a recent study [24] showing that the γ-phase mechanical stabilization is due to phonon-phonon interactions within a weakly correlated picture (DFT) of the electronic structure.Nevertheless, we apply FPLAPW (PAW) calculations for γ-uranium using the Wien2K (VASP) codes in order to explicitly explore the effects of LSDA + U. The results from the two implementations are in qualitative agreement with each other and summarized in Table 1. They predict a metastable non-magnetic (non-spin polarized, NSP) solution, that yield equilibrium V and B in better agreement with experiment than the spin polarized (SP) solution. The spin polarized LSDA + U ground-state solution (ferromagnetic, spin moment = 2.1 μB) has an unrealistically large equilibrium volume of 23 (23.6) Å3/atom, and substantially underestimates the bulk modulus. The total energy difference between the SP and the NSP states, at their respective equilibrium volumes, is a significant 64 (110) meV/atom. The finding of magnetism is contrary to the observed non-magnetic state of uranium metal. We thus conclude that the LSDA + U treatment, while essential for some aspects (i.e. electronic spectra) of the strongly correlated insulating uranium oxide compounds, is detrimental and incorrect for the more weakly correlated metallic U-Zr alloys.In conclusion we find that while the calculations by Landa et al. [2] can be mildly modified and improved by spin-orbit interaction, they do not suffer from fundamental inaccuracies as proposed in [1] and their enthalpy of formation of the bcc structure shall remain realistic.

Uranium contents in plants and mushrooms grown on a uranium-contaminated site near Ronneburg in Eastern Thuringia/Germany.

PubMed

Baumann, Nils; Arnold, Thuro; Haferburg, Götz

2014-01-01

Uranium concentrations in cultivated (sunflower, sunchoke, potato) and native plants, plant compartment specimens, and mushrooms, grown on a test site within a uranium-contaminated area in Eastern Thuringia, were analyzed and compared. This test site belongs to the Friedrich-Schiller University Jena and is situated on the ground of a former but now removed uranium mine waste leaching heap. For determination of the U concentrations in the biomaterials, the saps of the samples were squeezed out by using an ultracentrifuge, after that, the uranium concentrations in the saps and the remaining residue were measured, using ICP-MS. The study further showed that uranium concentrations observed in plant compartment and mushroom fruiting bodies sap samples were always higher than their associated solid residue sample. Also, it was found that the detected uranium concentration in the root samples were always higher than were observed in their associated above ground biomass, e.g., in shoots, leaves, blossoms etc. The highest uranium concentration was measured with almost 40 ppb U in a fruiting body of a mushroom and in roots of butterbur. However, the detected uranium concentrations in plants and mushrooms collected in this study were always lower than in the associated surface and soil water of the test site, indicating that under the encountered natural conditions, none of the studied plant and mushroom species turned out to be a hyperaccumulator for uranium, which could have extracted uranium in sufficient amounts out of the uranium-contaminated soil. In addition, it was found that the detected uranium concentrations in the sap samples, despite being above the sensitivity limit, proved to be too low-in combination with the presence of fluorescence quenching substances, e.g., iron and manganese ions, and/or organic quenchers-to extract a useful fluorescence signal, which could have helped to identify the uranium speciation in plants.

Uranium-Loaded Water Treatment Resins: 'Equivalent Feed' at NRC and Agreement State-Licensed Uranium Recovery Facilities - 12094

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

Camper, Larry W.; Michalak, Paul; Cohen, Stephen

Community Water Systems (CWSs) are required to remove uranium from drinking water to meet EPA standards. Similarly, mining operations are required to remove uranium from their dewatering discharges to meet permitted surface water discharge limits. Ion exchange (IX) is the primary treatment strategy used by these operations, which loads uranium onto resin beads. Presently, uranium-loaded resin from CWSs and mining operations can be disposed as a waste product or processed by NRC- or Agreement State-licensed uranium recovery facilities if that licensed facility has applied for and received permission to process 'alternate feed'. The disposal of uranium-loaded resin is costly andmore » the cost to amend a uranium recovery license to accept alternate feed can be a strong disincentive to commercial uranium recovery facilities. In response to this issue, the NRC issued a Regulatory Issue Summary (RIS) to clarify the agency's policy that uranium-loaded resin from CWSs and mining operations can be processed by NRC- or Agreement State-licensed uranium recovery facilities without the need for an alternate feed license amendment when these resins are essentially the same, chemically and physically, to resins that licensed uranium recovery facilities currently use (i.e., equivalent feed). NRC staff is clarifying its current alternate feed policy to declare IX resins as equivalent feed. This clarification is necessary to alleviate a regulatory and financial burden on facilities that filter uranium using IX resin, such as CWSs and mine dewatering operations. Disposing of those resins in a licensed facility could be 40 to 50 percent of the total operations and maintenance (O and M) cost for a CWS. Allowing uranium recovery facilities to treat these resins without requiring a license amendment lowers O and M costs and captures a valuable natural resource. (authors)« less

URANIUM IN ROCK MINERALS OF THE INTRUSION OF KYZL-OMPUL MOUNTAINS (NORTH KIRGISIA) (in Russian)

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

Leonova, L.L.; Pogiblova, L.S.

1961-01-01

The uranium distribution in rock minerals (syenites, granosyenites, and alaskite granites) of the Kyzyl-Ompul raassif is studied. Alaskite granites are characterized by the granite type of uranium distribution in minerals, about 50 percent of this element being connected with rockforming and about 50 percent with accessory uranium minerals. ln syenites uranium (about 70 percent) is bound to rockforming minerals. The same minerals from syenites and granites strongly differ by their uranium content and are constant in the ranges of each of those rock types. Granosyenites have aa intermediate (between syenites and granites) type of uranium distribution in minerals. (auth)

Removal of uranium from soil samples for ICP-OES analysis of RCRA metals

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

Wero, M.; Lederer-Cano, A.; Billy, C.

1995-12-01

Soil samples containing high levels of uranium present unique analytical problems when analyzed for toxic metals (Ag, As, Ba, Cd, Cr, Cu, Ni, Pb, Se and Tl) because of the spectral interference of uranium in the ICP-OES emission spectrometer. Methods to remove uranium from the digestates of soil samples, known to be high in uranium, have been developed that reduce the initial uranium concentration (1-3%) to less than 500 ppm. UTEVA ion exchange columns, used as an ICP-OES analytical pre-treatment, reduces uranium to acceptable levels, permitting good analytical results of the RCRA metals by ICP-OES.

PRODUCTION OF URANIUM

DOEpatents

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

1958-04-15

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

Depleted Uranium Program: Repository and Chemical Analysis of Biological Samples

DTIC Science & Technology

2010-11-01

Chemical Samples • Chemical Pathology and Analytical Assessment of U and DU in: • Tissues • Urine • Whole blood • Semen • Embedded fragments...preparation for determination of total uranium and isotopic uranium ratios  Semen – Total Uranium – dry ashed by concentrated nitric acid in muffle...Total uranium and DU measurements in blood 0.0 50.0 100.0 150.0 200.0 250.0 ng U in s am pl e Sample Number Semen Measured U Theortical U Uranium

Method of precipitating uranium from an aqueous solution and/or sediment

DOEpatents

Tokunaga, Tetsu K; Kim, Yongman; Wan, Jiamin

2013-08-20

A method for precipitating uranium from an aqueous solution and/or sediment comprising uranium and/or vanadium is presented. The method includes precipitating uranium as a uranyl vanadate through mixing an aqueous solution and/or sediment comprising uranium and/or vanadium and a solution comprising a monovalent or divalent cation to form the corresponding cation uranyl vanadate precipitate. The method also provides a pathway for extraction of uranium and vanadium from an aqueous solution and/or sediment.

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)

Rhizofiltration using sunflower (Helianthus annuus L.) and bean (Phaseolus vulgaris L. var. vulgaris) to remediate uranium contaminated groundwater.

PubMed

Lee, Minhee; Yang, Minjune

2010-01-15

The uranium removal efficiencies of rhizofiltration in the remediation of groundwater were investigated in lab-scale experiments. Sunflower (Helianthus annuus L.) and bean (Phaseolus vulgaris L. var. vulgaris) were cultivated and an artificially uranium contaminated solution and three genuine groundwater samples were used in the experiments. More than 80% of the initial uranium in solution and genuine groundwater, respectively, was removed within 24h by using sunflower and the residual uranium concentration of the treated water was lower than 30 microg/L (USEPA drinking water limit). For bean, the uranium removal efficiency of the rhizofiltration was roughly 60-80%. The maximum uranium removal via rhizofiltration for the two plant cultivars occurred at pH 3-5 of solution and their uranium removal efficiencies exceeded 90%. The lab-scale continuous rhizofiltration clean-up system delivered over 99% uranium removal efficiency, and the results of SEM and EDS analyses indicated that most uranium accumulated in the roots of plants. The present results suggested that the uranium removal capacity of two plants evaluated in the clean-up system was about 25mg/kg of wet plant mass. Notably, the removal capacity of the root parts only was more than 500 mg/kg.

Organic geochemical analysis of sedimentary organic matter associated with uranium

USGS Publications Warehouse

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

1986-01-01

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

Large decadal-scale changes in uranium and bicarbonate in groundwater of the irrigated western U.S

USGS Publications Warehouse

Burow, Karen R.; Belitz, Kenneth; Dubrovsky, Neil M.; Jurgens, Bryant C.

2017-01-01

Samples collected about one decade apart from 1105 wells from across the U.S. were compiled to assess whether uranium concentrations in the arid climate are linked to changing bicarbonate concentrations in the irrigated western U.S. Uranium concentrations in groundwater were high in the arid climate in the western U.S, where uranium sources are abundant. Sixty-four wells (6%) were above the U.S. EPA MCL of 30 μg/L; all but one are in the arid west. Concentrations were low to non-detectable in the humid climate. Large uranium and bicarbonate increases (differences are greater than the uncertainty in concentrations) occur in 109 wells between decade 1 and decade 2. Similarly, large uranium and bicarbonate decreases occur in 76 wells between the two decades. Significantly more wells are concordant (uranium and bicarbonate are both going the same direction) than discordant (uranium and bicarbonate are going opposite directions) (p < 0.001; Chi-square test). The largest percent difference in uranium concentrations occur in wells where uranium is increasing and bicarbonate is also increasing. These large differences occur mostly in the arid climate. Results are consistent with the hypothesis that changing uranium concentrations are linked to changes in bicarbonate in irrigated areas of the western U.S.

Immobilization of uranium into magnetite from aqueous solution by electrodepositing approach.

PubMed

Lu, Bing-Qing; Li, Mi; Zhang, Xiao-Wen; Huang, Chun-Mei; Wu, Xiao-Yan; Fang, Qi

2018-02-05

Immobilization of uranium into magnetite (Fe 3 O 4 ), which was generated from metallic iron by electrochemical method, was proposed to rapidly remove uranium from aqueous solution. The effects of electrochemical parameters such as electrode materials, voltage, electrode gap, reaction time and pH value on the crystallization of Fe 3 O 4 and uranium removal efficiencies were investigated. More than 90% uranium in the solution was precipitated with Fe 3 O 4 under laboratory conditions when uranium concentration range from 0.5mg/L to 10mg/L. The Fe 3 O 4 crystallization mechanism and immobilization of uranium was proved by XPS, XRD, TEM, FTIR and VSM methods. The results indicated that the cationic (including Fe 2+ , Fe 3+ and U(VI)) migrate to cathode side under the electric field and the uranium was incorporated or adsorbed by Fe 3 O 4 which was generated at cathode while the pH ranges between 2-7. The uranium-containing precipitate of Fe 3 O 4 can exist stably at the acid concentration below 60g/L. Furthermore, the precipitate may be used as valuable resources for uranium or iron recycling, which resulted in no secondary pollution in the removal of uranium from aqueous solution. Copyright © 2017. Published by Elsevier B.V.

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. 

Behavior of uranium under conditions of interaction of rocks and ores with subsurface water

NASA Astrophysics Data System (ADS)

Omel'Yanenko, B. I.; Petrov, V. A.; Poluektov, V. V.

2007-10-01

The behavior of uranium during interaction of subsurface water with crystalline rocks and uranium ores is considered in connection with the problem of safe underground insulation of spent nuclear fuel (SNF). Since subsurface water interacts with crystalline rocks formed at a high temperature, the mineral composition of these rocks and uranium species therein are thermodynamically unstable. Therefore, reactions directed toward the establishment of equilibrium proceed in the water-rock system. At great depths that are characterized by hindered water exchange, where subsurface water acquires near-neutral and reducing properties, the interaction is extremely sluggish and is expressed in the formation of micro- and nanoparticles of secondary minerals. Under such conditions, the slow diffusion redistribution of uranium with enrichment in absorbed forms relative to all other uranium species is realized as well. The products of secondary alteration of Fe- and Ti-bearing minerals serve as the main sorbents of uranium. The rate of alteration of minerals and conversion of uranium species into absorbed forms is slow, and the results of these processes are insignificant, so that the rocks and uranium species therein may be regarded as unaltered. Under reducing conditions, subsurface water is always saturated with uranium. Whether water interacts with rock or uranium ore, the equilibrium uranium concentration in water is only ≤10-8 mol/l. Uraninite ore under such conditions always remains stable irrespective of its age. The stability conditions of uranium ore are quite suitable for safe insulation of SNF, which consists of 95% uraninite (UO2) and is a confinement matrix for all other radionuclides. The disposal of SNF in massifs of crystalline rocks at depths below 500 m, where reducing conditions are predominant, is a reliable guarantee of high SNF stability. Under oxidizing conditions of the upper hydrodynamic zone, the rate of interaction of rocks with subsurface water increases by orders of magnitude and subsurface water is commonly undersaturated with uranium. Uranium absorbed by secondary minerals, particularly by iron hydroxides and leucoxene, is its single stable species under oxidizing conditions. The impact of oxygen-bearing water leads to destruction of uranium ore. This process is realized simultaneously at different hypsometric levels even if the permeability of the medium is variable in both the lateral and vertical directions. As a result, intervals containing uranyl minerals and relics of primary uranium ore are combined in ore-bearing zones with intervals of completely dissolved uranium minerals. A wide halo of elevated uranium contents caused by sorption is always retained at the location of uranium ore entirely destroyed by weathering. Uranium ore commonly finds itself in the aeration zone due to technogenic subsidence of the groundwater table caused by open-pit mining or pumping out of water from underground mines. The capillary and film waters that interact with rocks and ores in this zone are supplemented by free water filtering along fractures when rain falls or snow is thawing. The interaction of uranium ore with capillary water results in oxidation of uraninite, accompanied by loosening of the mineral surface, formation of microfractures, and an increase in solubility with enrichment of capillary water in uranium up to 10-4 mol/l. Secondary U(VI) minerals, first of all, uranyl hydroxides and silicates, replace uraninite, and uranium undergoes local diffusion redistribution with its sorption by secondary minerals of host rocks. The influx of free water facilitates the complete dissolution of primary and secondary uranium minerals, the removal of uranium at the sites of groundwater discharge, and its redeposition under reducing conditions at a greater depth. It is evident that the conditions of the upper hydrodynamic zone and the aeration zone are unfit for long-term insulation of SNF and high-level wastes because, after the failure of containers, the leakage of radionuclides into the environment becomes inevitable.

Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

NASA Astrophysics Data System (ADS)

Jue, Jan-Fong; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

2014-05-01

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U-Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U-10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are:

Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

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

Jan-Fong Jue; Dennis D. Keiser, Jr.; Cynthia R. Breckenridge

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Momore » fuel meat and Al–6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are • A typical Zr diffusion barrier of thickness 25 µm • Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 µm • Chemical banding, in some areas more than 100 µm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt% • Decomposed areas containing plate-shaped low-Mo phase • A typical Zr/cladding interaction layer of thickness 1-2 µm • A visible UZr2 bearing layer of thickness 1-2 µm • Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix • No excessive interaction between cladding and the uncoated fuel edge • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. • Some of these attributes might be critical to the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and interaction layer between U–Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.« less

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2011 CFR

2011-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2012 CFR

2012-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2014 CFR

2014-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2010 CFR

2010-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2013 CFR

2013-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

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.

PRODUCTION OF URANIUM MONOCARBIDE

DOEpatents

Powers, R.M.

1962-07-24

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

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

NASA Technical Reports Server (NTRS)

Roman, W. C.

1979-01-01

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

Preparation of carbon coated Fe3O4 nanoparticles for magnetic separation of uranium

NASA Astrophysics Data System (ADS)

Zhang, Xiaofei; Wang, Jun

2018-01-01

Uranium(VI) was removed from aqueous solutions using carbon coated Fe3O4 nanoparticles (Fe3O4@C). Batch experiments were conducted to study the effects of initial pH, shaking time and temperature on uranium sorption efficiency. It was found that the maximum adsorption capacity of the Fe3O4@C toward uranium(VI) was ∼120.20 mg g-1 when the initial uranium(VI) concentration was 100 mg L-1, displaying a high efficiency for the removal of uranium(VI) ions. Kinetics of the uranium(VI) removal is found to follow pseudo-second-order rate equation. In addition, the uranium(VI)-loaded Fe3O4@C nanoparticles can be recovered easily from aqueous solution by magnetic separation and regenerated by acid treatment. Present study suggested that magnetic Fe3O4@C composite particles can be used as an effective and recyclable adsorbent for the removal of uranium(VI) from aqueous solutions.

Can we predict uranium bioavailability based on soil parameters? Part 2: soil solution uranium concentration is not a good bioavailability index.

PubMed

Vandenhove, H; Van Hees, M; Wannijn, J; Wouters, K; Wang, L

2007-01-01

The present study aimed to quantify the influence of soil parameters on uranium uptake by ryegrass. Ryegrass was established on eighteen distinct soils, spiked with (238)U. Uranium soil-to-plant transfer factors (TF) ranged from 0.0003 to 0.0340kgkg(-1). There was no significant relation between the U soil-to-plant transfer (or total U uptake or flux) and the uranium concentration in the soil solution or any other soil factor measured, nor with the U recovered following selective soil extractions. Multiple linear regression analysis resulted in a significant though complex model explaining up to 99% of variation in TF. The influence of uranium speciation on uranium uptake observed was featured: UO(2)(+2), uranyl carbonate complexes and UO(2)PO(4)(-) seem the U species being preferentially taken up by the roots and transferred to the shoots. Improved correlations were obtained when relating the uranium TF with the summed soil solution concentrations of mentioned uranium species.

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

PubMed

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

2011-11-01

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

[Uranium exposure and cancer risk: a review of epidemiological studies].

PubMed

Tirmarche, M; Baysson, H; Telle-Lamberton, M

2004-02-01

At the end of 2000, certain diseases including leukemia were reported among soldiers who participated in the Balkan and in the Gulf wars. Depleted uranium used during these conflicts was considered as a possible cause. Its radiotoxicity is close to that of natural uranium. This paper reviews the epidemiological knowledge of uranium, the means of exposure and the associated risk of cancer. The only available epidemiological data concerns nuclear workers exposed to uranium. A review of the international literature is proposed by distinguishing between uranium miners and other workers of the nuclear industry. French studies are described in details. In ionizing radiation epidemiology, contamination by uranium is often cited as a risk factor, but the dose-effect relationship is rarely studied. Retrospective assessment of individual exposure is generally insufficient. Moreover, it is difficult to distinguish between uranium radiotoxicity, its chemical toxicity and the radiotoxicity of its progeny. A causal relation between lung cancer and radon exposure, a gas derived from the decay of uranium, has been demonstrated in epidemiological studies of miners. Among other nuclear workers exposed to uranium, there is a mortality deficit from all causes (healthy worker effect). No cancer site appears systematically in excess compared to the national population; very few studies describe a dose-response relationship. Only studies with a precise reconstruction of doses and sufficient numbers of workers will allow a better assessment of risks associated with uranium exposure at levels encountered in industry or during conflicts using depleted uranium weapons.

Enhanced uranium immobilization and reduction by Geobacter sulfurreducens biofilms.

PubMed

Cologgi, Dena L; Speers, Allison M; Bullard, Blair A; Kelly, Shelly D; Reguera, Gemma

2014-11-01

Biofilms formed by dissimilatory metal reducers are of interest to develop permeable biobarriers for the immobilization of soluble contaminants such as uranium. Here we show that biofilms of the model uranium-reducing bacterium Geobacter sulfurreducens immobilized substantially more U(VI) than planktonic cells and did so for longer periods of time, reductively precipitating it to a mononuclear U(IV) phase involving carbon ligands. The biofilms also tolerated high and otherwise toxic concentrations (up to 5 mM) of uranium, consistent with a respiratory strategy that also protected the cells from uranium toxicity. The enhanced ability of the biofilms to immobilize uranium correlated only partially with the biofilm biomass and thickness and depended greatly on the area of the biofilm exposed to the soluble contaminant. In contrast, uranium reduction depended on the expression of Geobacter conductive pili and, to a lesser extent, on the presence of the c cytochrome OmcZ in the biofilm matrix. The results support a model in which the electroactive biofilm matrix immobilizes and reduces the uranium in the top stratum. This mechanism prevents the permeation and mineralization of uranium in the cell envelope, thereby preserving essential cellular functions and enhancing the catalytic capacity of Geobacter cells to reduce uranium. Hence, the biofilms provide cells with a physically and chemically protected environment for the sustained immobilization and reduction of uranium that is of interest for the development of improved strategies for the in situ bioremediation of environments impacted by uranium contamination. Copyright © 2014, American Society for Microbiology. All Rights Reserved.

Enhanced Uranium Immobilization and Reduction by Geobacter sulfurreducens Biofilms

PubMed Central

Cologgi, Dena L.; Speers, Allison M.; Bullard, Blair A.; Kelly, Shelly D.

2014-01-01

Biofilms formed by dissimilatory metal reducers are of interest to develop permeable biobarriers for the immobilization of soluble contaminants such as uranium. Here we show that biofilms of the model uranium-reducing bacterium Geobacter sulfurreducens immobilized substantially more U(VI) than planktonic cells and did so for longer periods of time, reductively precipitating it to a mononuclear U(IV) phase involving carbon ligands. The biofilms also tolerated high and otherwise toxic concentrations (up to 5 mM) of uranium, consistent with a respiratory strategy that also protected the cells from uranium toxicity. The enhanced ability of the biofilms to immobilize uranium correlated only partially with the biofilm biomass and thickness and depended greatly on the area of the biofilm exposed to the soluble contaminant. In contrast, uranium reduction depended on the expression of Geobacter conductive pili and, to a lesser extent, on the presence of the c cytochrome OmcZ in the biofilm matrix. The results support a model in which the electroactive biofilm matrix immobilizes and reduces the uranium in the top stratum. This mechanism prevents the permeation and mineralization of uranium in the cell envelope, thereby preserving essential cellular functions and enhancing the catalytic capacity of Geobacter cells to reduce uranium. Hence, the biofilms provide cells with a physically and chemically protected environment for the sustained immobilization and reduction of uranium that is of interest for the development of improved strategies for the in situ bioremediation of environments impacted by uranium contamination. PMID:25128347

Regularities of spatial association of major endogenous uranium deposits and kimberlitic dykes in the uranium ore regions of the Ukrainian Shield

NASA Astrophysics Data System (ADS)

Kalashnyk, Anna

2015-04-01

During exploration works we discovered the spatial association and proximity time formation of kimberlite dykes (ages are 1,815 and 1,900 Ga for phlogopite) and major industrial uranium deposits in carbonate-sodium metasomatites (age of the main uranium ore of an albititic formation is 1,85-1,70 Ga according to U-Pb method) in Kirovogradsky, Krivorozhsky and Alekseevsko-Lysogorskiy uranium ore regions of the Ukrainian Shield (UkrSh) [1]. In kimberlites of Kirovogradsky ore region uranium content reaches 18-20 g/t. Carbon dioxide is a major component in the formation of hydrothermal uranium deposits and the formation of the sodium in the process of generating the spectrum of alkaline ultrabasic magmas in the range from picritic to kimberlite and this is the connection between these disparate geochemical processes. For industrial uranium deposits in carbonate-sodium metasomatitics of the Kirovogradsky and Krivorozhsky uranium ore regions are characteristic of uranyl carbonate introduction of uranium, which causes correlation between CO2 content and U in range of "poor - ordinary - rich" uranium ore. In productive areas of uranium-ore fields of the Kirovogradsky ore region for phlogopite-carbonate veinlets of uranium ore albitites deep δ13C values (from -7.9 to -6.9o/oo) are characteristic. Isotope-geochemical investigation of albitites from Novokonstantynovskoe, Dokuchaevskoe, Partyzanskoe uranium deposits allowed obtaining direct evidence of the involvement of mantle material during formation of uranium albitites in Kirovogradsky ore region [2]. Petrological characteristics of kimberlites from uranium ore regions of the UkrSh (presence of nodules of dunite and harzburgite garnet in kimberlites, diamonds of peridotite paragenesis, chemical composition of indicator minerals of kimberlite, in particular Gruzskoy areas pyropes (Cr2O3 = 6,1-7,1%, MgO = 19,33-20,01%, CaO = 4,14-4,38 %, the content of knorringite component of most grains > 50mol%), chromites (Cr2O3 = 45,32-62,17%, MgO = 7,3-12,5%) allow us to estimate the depth of generation of kimberlite magmas more than 170-200 km. Ilmenites show two groups according to MgO, Cr2O3 and TiO2 content. Reconstructions of the mantle sections show also two intervals of pressures divided at 4.5 GPa, the upper part is highly metasomatized This high degree metasomatism is determined for almost all mantle columns. It is suggested that large-scale of uranium-bearing mantle fluids may be associated with the ancient degasation during the subduction which is highly enriched in U component . Analysis of the reasons for the marked association kimberlitic dykes and major industrial uranium deposits in carbonate-sodium metasomatic in the UkrSh led to the conclusion that hydrothermal uranium deposits are confined to the supply mantle fluid systems of mantle fault zones exercising brings sodium carbonate solutions enriched uranium from mantle sources. References: 1. Kalashnik A.A. New prognostic-evaluation criteria in technology prognosis of forming industrial endogenous uranium deposits of the Ukrainian Shield, 2014. Scientific proceedings of UkrSGRI, № 2, p. 27-54 (in Russian) 2. Stepanjuk L.M., Bondarenko S.V., Somka V.O. and other, 2012. Source of uranium and uranium-bearing sodium albitites for example of Dokuchaievskogo field of the Ingulsky megablock of the UkrSh: Abstracts of scientific conference "Theoretical issues and research practice metasomatic rocks and ores" (Kyiv, 14-16 March 2012), IGMOF, p.78-80. (in Ukrainian)

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.

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.

Exposure assessment of natural uranium from drinking water.

PubMed

Jakhu, Rajan; Mehra, Rohit; Mittal, H M

2016-12-08

The uranium concentration in the drinking water of the residents of the Jaipur and Ajmer districts of Rajasthan has been measured for exposure assessment. The daily intake of uranium from the drinking water for the residents of the study area is found to vary from 0.4 to 123.9 μg per day. For the average uranium ingestion rate of 35.2 μg per day for a long term exposure period of 60 years, estimations have been made for the retention of uranium in different body organs and its excretion with time using ICRP's biokinetic model of uranium. Radioactive and chemical toxicity of uranium has been reported and discussed in detail in the present manuscript.

METHOD OF APPLYING COPPER COATINGS TO URANIUM

DOEpatents

Gray, A.G.

1959-07-14

A method is presented for protecting metallic uranium, which comprises anodic etching of the uranium in an aqueous phosphoric acid solution containing chloride ions, cleaning the etched uranium in aqueous nitric acid solution, promptly electro-plating the cleaned uranium in a copper electro-plating bath, and then electro-plating thereupon lead, tin, zinc, cadmium, chromium or nickel from an aqueous electro-plating bath.

Reconnaissance for uranium in asphalt-bearing rocks in the western states

USGS Publications Warehouse

Hail, William James

1955-01-01

Evaluation of field data indicates that naturally occurring asphalts with a relatively high uranium content probably originated in, or migrated through, rocks that contain more than average amounts of uranium. It is believed that some of the uranium was present as an original constituent of the oil but that some uranium may have been introduced during migration of the oil.

DISSOLUTION OF URANIUM FUELS BY MONOOR DIFLUOROPHOSPHORIC ACID

DOEpatents

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

1963-05-01

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