National Library of Energy BETA

Sample records for domestic uranium production

  1. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Number of Holes Feet (thousand) Number of Holes ...

  2. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Table 9. Summary production statistics of the U.S. ...

  3. 2015 Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Domestic Uranium Production Report 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 State(s) 2003 2004 2005 2006 2007 2008 2009 2010 ...

  4. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    5 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Production Mining Method 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 ...

  5. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    9. Summary production statistics of the U.S. uranium industry, 1993-2015 Year Exploration and development surface drilling (million feet) Exploration and development drilling expenditures 1 (million dollars) Mine production of uranium (million pounds U3O8) Uranium concentrate production (million pounds U3O8) Uranium concentrate shipments (million pounds U3O8) Employment (person-years) 1993 1.1 5.7 2.1 3.1 3.4 871 1994 0.7 1.1 2.5 3.4 6.3 980 1995 1.3 2.6 3.5 6.0 5.5 1,107 1996 3.0 7.2 4.7 6.3

  6. 2015 Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    7 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Capacity (short tons of ore per day) 2011 2012 2013 2014 2015 Anfield Resources ...

  7. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    9 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Year Exploration Mining Milling Processing Reclamation Total 2003 W W W W 117 321 ...

  8. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    10. Uranium reserve estimates at the end of 2014 and 2015 million pounds U3O8 End of 2014 End of 2015 Forward Cost2 Uranium Reserve Estimates1 by Mine and Property Status, Mining Method, and State(s) $0 to $30 per pound $0 to $50 per pound $0 to $100 per pound $0 to $30 per pound $0 to $50 per pound $0 to $100 per pound Properties with Exploration Completed, Exploration Continuing, and Only Assessment Work W W 154.6 24.3 W 151.6 Properties Under Development for Production and Development

  9. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    3. U.S. uranium concentrate production, shipments, and sales, 2003-15 Activity at U.S. mills and In-Situ-Leach plants 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015 Estimated contained U3O8 (thousand pounds) Ore from Mines and Stockpiles Fed to Mills1 0 W W W 0 W W W W W W W 0 Other Feed Materials 2 W W W W W W W W W W W W W Total Mill Feed W W W W W W W W W W W W W Uranium Concentrate Produced at U.S. Mills (thousand pounds U3O8) W W W W W W W W W W W W W Uranium Concentrate

  10. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    and Development Drilling","Mine Production of Uranium ","Uranium Concentrate Production ","Uranium Concentrate Shipments ","Employment " "Year","Drilling (million feet)"," ...

  11. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    2. U.S. uranium mine production and number of mines and sources, 2003-15 Production / Mining method 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015 Underground (estimated contained thousand pounds U3O8) W W W W W W W W W W W W W Open Pit (estimated contained thousand pounds U3O8) 0 0 0 0 0 0 0 0 0 0 0 0 0 In-Situ Leaching (thousand pounds U3O8) W W 2,681 4,259 W W W W W W W W W Other1 (thousand pounds U3O8) W W W W W W W W W W W W W Total Mine Production (thousand pounds U3O8)

  12. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    6. Employment in the U.S. uranium production industry by category, 2003-15 person-years Year Exploration Mining Milling Processing Reclamation Total 2003 W W W W 117 321 2004 18 108 W W 121 420 2005 79 149 142 154 124 648 2006 188 121 W W 155 755 2007 375 378 107 216 155 1,231 2008 457 558 W W 154 1,563 2009 175 441 W W 162 1,096 2010 211 400 W W 125 1,073 2011 208 462 W W 102 1,191 2012 161 462 W W 179 1,196 2013 149 392 W W 199 1,156 2014 86 246 W W 161 787 2015 58 251 W W 116

  13. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    7. Employment in the U.S. uranium production industry by state, 2003-15 person-years State(s) 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015 Wyoming 134 139 181 195 245 301 308 348 424 512 531 416 343 Colorado and Texas 48 140 269 263 557 696 340 292 331 248 198 105 79 Nebraska and New Mexico 92 102 123 160 149 160 159 134 127 W W W W Arizona, Utah, and Washington 47 40 75 120 245 360 273 281 W W W W W Alaska, Michigan, Nevada, and South Dakota 0 0 0 16 25 30 W W W W W 0 0

  14. Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Resources, Inc., dba Cameco Resources Smith Ranch-Highland Operation Converse, Wyoming ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  15. 2015 Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    11 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Total Land and Other 2003 W W 31.3 NA NA NA W 2004 10.6 27.8 48.4 NA NA NA 86.9 ...

  16. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    8. U.S. uranium expenditures, 2003-15 million dollars Year Drilling1 Production2 Land and other 3 Total expenditures Total land and other Land Exploration Reclamation 2003 W W 31.3 NA NA NA W 2004 10.6 27.8 48.4 NA NA NA 86.9 2005 18.1 58.2 59.7 NA NA NA 136.0 2006 40.1 65.9 115.2 41.0 23.3 50.9 221.2 2007 67.5 90.4 178.2 77.7 50.3 50.2 336.2 2008 81.9 221.2 164.4 65.2 50.2 49.1 467.6 2009 35.4 141.0 104.0 17.3 24.2 62.4 280.5 2010 44.6 133.3 99.5 20.2 34.5 44.7 277.3 2011 53.6 168.8 96.8 19.6

  17. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    4. U.S. uranium mills by owner, location, capacity, and operating status at end of the year, 2011-15 Owner Mill and Heap Leach1 Facility name County, state (existing and planned locations) Capacity (short tons of ore per day) Operating status at end of the year 2011 2012 2013 2014 2015 Anfield Resources Shootaring Canyon Uranium Mill Garfield, Utah 750 Standby Standby Standby Standby Standby EPR White Mesa LLC White Mesa Mill San Juan, Utah 2,000 Operating Operating Operating- Processing

  18. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    1. U.S. uranium drilling activities, 2003-15 Exploration drilling Development drilling Exploration and development drilling Year Number of holes Feet (thousand) Number of holes Feet (thousand) Number of holes Feet (thousand) 2003 NA NA NA NA W W 2004 W W W W 2,185 1,249 2005 W W W W 3,143 1,668 2006 1,473 821 3,430 1,892 4,903 2,713 2007 4,351 2,200 4,996 2,946 9,347 5,146 2008 5,198 2,543 4,157 2,551 9,355 5,093 2009 1,790 1,051 3,889 2,691 5,679 3,742 2010 2,439 1,460 4,770 3,444 7,209 4,904

  19. Domestic Uranium Production Report - Energy Information Administration

    U.S. Energy Information Administration (EIA) Indexed Site

    Domestic Uranium Production Report - Annual With Data for 2015 | Release Date: May 5, 2016 | Next Release Date: May 2017 | full report Previous domestic uranium production reports Year: 2014 2013 2012 2011 2010 2009 2008 2007 2006 2005 2004 Go Drilling Total uranium drilling was 1,518 holes covering 0.9 million feet, 13% fewer holes than in 2015. Expenditures for uranium drilling in the United States were $29 million in 2015, an increase of 2% compared with 2014. Figure 1. U.S. Uranium drilling

  20. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Forward costs are neither the full costs of production nor the market price at which the uranium, when produced, might be sold." "Note: Totals may not equal sum of components ...

  1. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    7. Employment in the U.S. uranium production industry by state, 2003-15" "person-years" "State(s)",2003,2004,2005,2006,2007,2008,2009,2010,2011,2012,2013,2014,2015 ...

  2. 1st Quarter 2016 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. Energy Information Administration: Form EIA-851A and Form EIA-851Q, ""Domestic Uranium Production Report.""" " U.S. Energy Information Administration Domestic Uranium ...

  3. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    data set of uranium reserves that were published in the July 2010 report U.S. Uranium Reserves Estimates at http:www.eia.govcneafnuclearpagereservesures.html. ...

  4. Domestic Uranium Production Report - Quarterly

    Gasoline and Diesel Fuel Update (EIA)

    1. Total production of uranium concentrate in the United States, 1996 - 2nd Quarter 2016 pounds U3O8 Calendar-year quarter 1st quarter 2nd quarter 3rd quarter 4th quarter Calendar-year total 1996 1,734,427 1,460,058 1,691,796 1,434,425 6,320,706 1997 1,149,050 1,321,079 1,631,384 1,541,052 5,642,565 1998 1,151,587 1,143,942 1,203,042 1,206,003 4,704,574 1999 1,196,225 1,132,566 1,204,984 1,076,897 4,610,672 2000 1,018,683 983,330 981,948 973,585 3,975,545 2001 709,177 748,298 628,720 553,060

  5. 2014 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Power Resources Inc., dba Cameco Resources Smith Ranch-Highland Operation Converse, ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  6. Domestic Uranium Production Report - Quarterly

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Resources, Inc. dba Cameco Resources Smith Ranch-Highland Operation Converse, Wyoming ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  7. 2014 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Resources Inc., dba Cameco Resources","Smith Ranch-Highland Operation","Converse, ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  8. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Activity at U.S. Mills and In-Situ-Leach Plants 2003 2004 2005 2006 2007 2008 2009 2010 ... Total Uranium Concentrate Shipped from U.S. Mills and In-Situ-Leach Plants Table 3. U.S. ...

  9. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Uranium Concentrate Sales by U.S. Producers 3" "Deliveries (thousand pounds U3O8)","W","W","W",3786,3602,3656,2044,2684,2870,3630,4447,4746,3634 "Weighted-Average Price ...

  10. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    May 5, 2016" "Next Release Date: May 2017" "Table 4. U.S. uranium mills and heap leach facilities by owner, location, capacity, and operating status at end of the year, 2011-15" ...

  11. Domestic Uranium Production Report - Quarterly

    Gasoline and Diesel Fuel Update (EIA)

    2. Number of uranium mills and plants producing uranium concentrate in the United States Uranium concentrate processing facilities End of Mills - conventional milling 1 Mills - other operations 2 In-situ-leach plants 3 Byproduct recovery plants 4 Total 1996 0 2 5 2 9 1997 0 3 6 2 11 1998 0 2 6 1 9 1999 1 2 4 0 7 2000 1 2 3 0 6 2001 0 1 3 0 4 2002 0 1 2 0 3 2003 0 0 2 0 2 2004 0 0 3 0 3 2005 0 1 3 0 4 2006 0 1 5 0 6 2007 0 1 5 0 6 2008 1 0 6 0 7 2009 0 1 3 0 4 2010 1 0 4 0 5 2011 1 0 5 0 6 2012 1

  12. Domestic Uranium Production Report - Quarterly

    Gasoline and Diesel Fuel Update (EIA)

    3. U.S. uranium mills and heap leach facilities by owner, location, capacity, and operating status Operating status at the end of Owner Mill and Heap Leach1 Facility name County, state (existing and planned locations) Capacity (short tons of ore per day) 2015 1st Quarter 2016 2nd quarter 2016 Anfield Resources Inc. Shootaring Canyon Uranium Mill Garfield, Utah 750 Standby Standby Standby EFR White Mesa LLC White Mesa Mill San Juan, Utah 2,000 Operating-Processing Alternate Feed

  13. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    mines and sources, 2003-15" "Production Mining Method",2003,2004,2005,2006,2007,2008,200... "Total Mines and Sources",4,6,10,11,12,17,20,9,11,12,12,11,9 "E Estimated data. ...

  14. Domestic Uranium Production Report - Quarterly - Energy Information...

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. uranium in-situ-leach plants in production (state) Crow Butte Operation (Nebraska) Lost Creek Project (Wyoming) Nichols Ranch ISR Project (Wyoming) Ross CPP (Wyoming) Smith ...

  15. 1st Quarter 2016 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Table 1. Total production of uranium concentrate in the United States, 1996 - 1st quarter 2016 pounds U 3 O 8 Calendar-year quarter 1st quarter 2nd quarter 3rd quarter 4th quarter ...

  16. 1st Quarter 2016 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Figure 1. Uranium concentrate production in the United States, 1996 - 1st quarter 2016 pounds U 3 O 8 0 500,000 1,000,000 1,500,000 2,000,000 2,500,000 3,000,000 3,500,000 ...

  17. 2nd Quarter 2016 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    2. Number of uranium mills and plants producing uranium concentrate in the United States" ... - other operations 2","In-situ-leach plants 3","Byproduct recovery plants 4","Total" ...

  18. 1st Quarter 2016 Domestic Uranium Production Report

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Table 2. Number of uranium mills and plants producing uranium concentrate in the United States End of Mills - conventional milling 1 Mills - other operations 2 In-situ-leach plants ...

  19. 2nd Quarter 2016 Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Power Resources, Inc. dba Cameco Resources Smith Ranch-Highland Operation Converse, ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  20. 2nd Quarter 2016 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Resources, Inc. dba Cameco Resources","Smith Ranch-Highland Operation","Converse, ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  1. 1st Quarter 2016 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Capacity (short tons of ore per day) 2015 1st quarter 2016 Anfield Resources Inc. Shootaring Canyon Uranium Mill Garfield, Utah 750 Standby Standby EFR White Mesa LLC White Mesa ...

  2. 1st Quarter 2016 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    3. U.S. uranium mills and heap leach facilities by owner, location, capacity, and operating status" "Owner","Mill and Heap Leach1 Facility name","County, state (existing and ...

  3. PRODUCTION OF URANIUM TETRACHLORIDE

    DOE Patents [OSTI]

    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.

  4. PRODUCTION OF PURIFIED URANIUM

    DOE Patents [OSTI]

    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.

  5. Crude Oil Domestic Production

    U.S. Energy Information Administration (EIA) Indexed Site

    Data Series: Crude Oil Domestic Production Refinery Crude Oil Inputs Refinery Gross Inputs Refinery Operable Capacity (Calendar Day) Refinery Percent Operable Utilization Net Inputs of Motor Gasoline Blending Components Net Inputs of RBOB Blending Components Net Inputs of CBOB Blending Components Net Inputs of GTAB Blending Components Net Inputs of All Other Blending Components Net Inputs of Fuel Ethanol Net Production - Finished Motor Gasoline Net Production - Finished Motor Gasoline (Excl.

  6. PRODUCTION OF URANIUM

    DOE Patents [OSTI]

    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.

  7. Domestic production of medical isotope Mo-99 moves a step closer

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Domestic production of medical isotope Mo-99 Domestic production of medical isotope Mo-99 moves a step closer Irradiated uranium fuel has been recycled and reused for molybdenum-99 ...

  8. Uranium in US surface, ground, and domestic waters. Volume 2

    SciTech Connect (OSTI)

    Drury, J.S.; Reynolds, S.; Owen, P.T.; Ross, R.H.; Ensminger, J.T.

    1981-04-01

    The report Uranium in US Surface, Ground, and Domestic Waters comprises four volumes. Volumes 2, 3, and 4 contain data characterizing the location, sampling date, type, use, and uranium conentrations of 89,994 individual samples presented in tabular form. The tabular data in volumes 2, 3, and 4 are summarized in volume 1 in narrative form and with maps and histograms.

  9. Domestic Uranium Production Report - Quarterly

    Gasoline and Diesel Fuel Update (EIA)

    Thank

  10. 2015 Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    ... Nichols Ranch ISR Project, Smith Ranch-Highland Operation, and Willow Creek Project). ... Ranch ISR Project, Smith Ranch-Highland Operation, and Willow Creek Project in Wyoming). ...

  11. PRODUCTION OF URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Fowler, R.D.

    1957-08-27

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

  12. State Support of Domestic Production

    SciTech Connect (OSTI)

    Amy Wright

    2007-12-30

    This project was developed in response to a cooperative agreement offering by the U.S. Department of Energy (DOE) and the National Energy Technology Laboratory (NETL) under the State Support of Domestic Production DE-FC26-04NT15456. The Interstate Oil and Gas Compact Commission (IOGCC) performed efforts in support of State programs related to the security, reliability and growth if our nation's domestic production of oil and natural gas. The project objectives were to improve the States ability to monitor the security of oil and gas operations; to maximize the production of domestic oil and natural gas thereby minimizing the threat to national security posed by interruptions in energy imports; to assist States in developing and maintaining high standards of environmental protection; to assist in addressing issues that limit the capacity of the industry; to promote the deployment of the appropriate application of technology for regulatory efficiency; and to inform the public about emerging energy issues.

  13. PRODUCTION OF URANIUM MONOCARBIDE

    DOE Patents [OSTI]

    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)

  14. Domestic production of medical isotope Mo-99 moves a step closer

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Domestic Uranium Production Report - Annual With Data for 2015 | Release Date: May 5, 2016 | Next Release Date: May 2017 | full report Previous domestic uranium production reports Year: 2014 2013 2012 2011 2010 2009 2008 2007 2006 2005 2004 Go Drilling Total uranium drilling was 1,518 holes covering 0.9 million feet, 13% fewer holes than in 2015. Expenditures for uranium drilling in the United States were $29 million in 2015, an increase of 2% compared with 2014. Figure 1. U.S. Uranium drilling

  15. PRODUCTION OF URANIUM TETRAFLUORIDE

    DOE Patents [OSTI]

    Shaw, W.E.; Spenceley, R.M.; Teetzel, F.M.

    1959-08-01

    A method is presented for producing uranium tetrafluoride from the gaseous hexafluoride by feeding the hexafluoride into a high temperature zone obtained by the recombination of molecularly dissociated hydrogen. The molal ratio of hydrogen to uranium hexnfluoride is preferably about 3 to 1. Uranium tetrafluoride is obtained in a finely divided, anhydrous state.

  16. ELECTROLYTIC PRODUCTION OF URANIUM TETRAFLUORIDE

    DOE Patents [OSTI]

    Lofthouse, E.

    1954-08-31

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

  17. PRODUCTION OF URANIUM

    DOE Patents [OSTI]

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

    1957-11-12

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

  18. Process for continuous production of metallic uranium and uranium alloys

    DOE Patents [OSTI]

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

    1995-06-06

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

  19. Process for continuous production of metallic uranium and uranium alloys

    DOE Patents [OSTI]

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

    1995-01-01

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

  20. PROCESS FOR PRODUCTION OF URANIUM

    DOE Patents [OSTI]

    Crawford, J.W.C.

    1959-09-29

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

  1. PROCESS FOR SEPARATING URANIUM FISSION PRODUCTS

    DOE Patents [OSTI]

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

    1959-03-10

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

  2. Uranium

    SciTech Connect (OSTI)

    Gabelman, J.W.; Chenoweth, W.L.; Ingerson, E.

    1981-10-01

    The uranium production industry is well into its third recession during the nuclear era (since 1945). Exploration is drastically curtailed, and many staffs are being reduced. Historical market price production trends are discussed. A total of 3.07 million acres of land was acquired for exploration; drastic decrease. Surface drilling footage was reduced sharply; an estimated 250 drill rigs were used by the uranium industry during 1980. Land acquisition costs increased 8%. The domestic reserve changes are detailed by cause: exploration, re-evaluation, or production. Two significant discoveries of deposits were made in Mohave County, Arizona. Uranium production during 1980 was 21,850 short tons U/sub 3/O/sub 8/; an increase of 17% from 1979. Domestic and foreign exploration highlights were given. Major producing areas for the US are San Juan basin, Wyoming basins, Texas coastal plain, Paradox basin, northeastern Washington, Henry Mountains, Utah, central Colorado, and the McDermitt caldera in Nevada and Oregon. 3 figures, 8 tables. (DP)

  3. Domestic Uranium Production Report - Energy Information Administration

    U.S. Energy Information Administration (EIA) Indexed Site

    Nichols Ranch ISR Project, Smith Ranch-Highland Operation, and Willow Creek Project). ... Ranch ISR Project, Smith Ranch-Highland Operation, and Willow Creek Project in Wyoming). ...

  4. Domestic Uranium Production Report 2004-13

    Gasoline and Diesel Fuel Update (EIA)

    Thank

  5. Domestic Uranium Production Report - Quarterly - Energy Information...

    U.S. Energy Information Administration (EIA) Indexed Site

    Crow Butte Operation (Nebraska) Lost Creek Project (Wyoming) Nichols Ranch ISR Project (Wyoming) Ross CPP (Wyoming) Smith Ranch-Highland Operation (Wyoming) Willow Creek Project ...

  6. President Truman Increases Production of Uranium and Plutonium...

    National Nuclear Security Administration (NNSA)

    Increases Production of Uranium and Plutonium President Truman Increases Production of Uranium and Plutonium Washington, DC President Truman approves a 1.4 billion expansion of ...

  7. SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS

    DOE Patents [OSTI]

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

    1959-10-13

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

  8. Decommissioning of U.S. Uranium Production Facilities

    Reports and Publications (EIA)

    1995-01-01

    This report analyzes the uranium production facility decommissioning process and its potential impact on uranium supply and prices. 1995 represents the most recent publication year.

  9. Uranium Industry Annual, 1992

    SciTech Connect (OSTI)

    Not Available

    1993-10-28

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ``Decommissioning of US Conventional Uranium Production Centers,`` is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2.

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

    DOE Patents [OSTI]

    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)

  11. Assessment of Defense-Related Uranium Mines in the United States and Other U.S. Department of Energy Office of Legacy Management (LM) Domestic and International Uranium Initiatives – 15335

    SciTech Connect (OSTI)

    Edge, Russel; Butherus, Michael; Ford, John

    2015-03-01

    Assessment of Defense-Related Uranium Mines in the United States and Other U.S. Department of Energy Office of Legacy Management (LM) Domestic and International Uranium Initiatives – 15335

  12. EA-1929: NorthStar Medical Technologies LLC, Commercial Domestic Production of the Medical Isotope Molybdenum-99

    Broader source: Energy.gov [DOE]

    This EA evaluates the potential environmental impacts of a proposal to use federal funds to support and accelerate Northstar Medical Radioisotopes' project to develop domestic, commercial production capability for the medical isotope Molybdenum-99 without the use of highly enriched uranium.

  13. Summary history of domestic uranium procurement under US Atomic Energy Commission contracts. Final report

    SciTech Connect (OSTI)

    Albrethsen, H. Jr.; McGinley, F.E.

    1982-09-01

    During the period 1947 through 1970, the Atomic Energy Commission (AEC) fostered the rapid development and expansion of the domestic uranium mining and milling industry by providing a market for uranium. Some thirty-two mills were constructed during that period to produce U/sub 3/O/sub 8/ concentrates for sale to the AEC. In addition, there were various pilot plants, concentrators, upgraders, heap leach, and solution mining facilities that operated during the period. The purpose of this report is to compile a short narrative history of the AEC's uranium concentrate procurement program and to describe briefly each of the operations that produced uranium for sale to the AEC. Contractual arrangements are described and data are given on quantities of U/sub 3/O/sub 8/ purchased and prices paid. Similar data are included for V/sub 2/O/sub 5/, where applicable. Mill and other plant operating data were also compiled from old AEC records. These latter data were provided by the companies, as a contractual requirement, during the period of operation under AEC contracts. Additionally, an effort was made to determine the present status of each facility by reference to other recently published reports. No sites were visited nor were the individual reports reviewed by the companies, many of which no longer exist. The authors relied almost entirely on published information for descriptions of facilities and milling processes utilized.

  14. President Truman Increases Production of Uranium and Plutonium | National

    National Nuclear Security Administration (NNSA)

    Nuclear Security Administration | (NNSA) Increases Production of Uranium and Plutonium President Truman Increases Production of Uranium and Plutonium Washington, DC President Truman approves a $1.4 billion expansion of Atomic Energy Commission facilities to produce uranium and plutonium for nuclear weapons

  15. PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

    DOE Patents [OSTI]

    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.

  16. Table 5. Domestic Crude Oil Production, Projected vs. Actual

    U.S. Energy Information Administration (EIA) Indexed Site

    Domestic Crude Oil Production, Projected vs. Actual" "Projected" " (million barrels)" ,1993,1994,1995,1996,1997,1998,1999,2000,2001,2002,2003,2004,2005,2006,2007,2008,2009,2010,201...

  17. final ERI-2142 18-1501 Analysis of Potential Effects on Domestic Industries of DOE Excess Uranium Inventory 2015-2024.docx

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    ERI-2142.18-1501 Analysis of the Potential Effects on the Domestic Uranium Mining, Conversion and Enrichment Industries of the Introduction of DOE Excess Uranium Inventory During CY 2015 Through 2024 ENERGY RESOURCES INTERNATIONAL, INC. 1015 18 th Street, NW, Suite 650 Washington, DC 20036 USA Telephone: (202) 785-8833 Facsimile: (202) 785-8834 ERI-2142.18-1501 Analysis of the Potential Effects on the Domestic Uranium Mining, Conversion and Enrichment Industries of the Introduction of DOE

  18. The Electrolytic Production of Metallic Uranium

    DOE Patents [OSTI]

    Rosen, R.

    1950-08-22

    This patent covers a process for producing metallic uranium by electrolyzing uranium tetrafluoride at an elevated temperature in a fused bath consisting essentially of mixed alkali and alkaline earth halides.

  19. Development of uranium metal targets for {sup 99}Mo production

    SciTech Connect (OSTI)

    Wiencek, T.C.; Hofman, G.L.

    1993-10-01

    A substantial amount of high enriched uranium (HEU) is used for the production of medical-grade {sup 99}Mo. Promising methods of producing irradiation targets are being developed and may lead to the reduction or elimination of this HEU use. To substitute low enriched uranium (LEU) for HEU in the production of {sup 99}Mo, the target material may be changed to uranium metal foil. Methods of fabrication are being developed to simplify assembly and disassembly of the targets. Removal of the uranium foil after irradiation without dissolution of the cladding is a primary goal in order to reduce the amount of liquid radioactive waste material produced in the process. Proof-of-concept targets have been fabricated. Destructive testing indicates that acceptable contact between the uranium foil and the cladding can be achieved. Thermal annealing tests, which simulate the cladding/uranium diffusion conditions during irradiation, are underway. Plans are being made to irradiate test targets.

  20. DOE/NNSA Successfully Establishes Uranium Lease and Takeback...

    National Nuclear Security Administration (NNSA)

    the DOENNSA's ongoing support for the establishment of a domestic, commercial Mo-99 production capability that does not use proliferation-sensitive highly enriched uranium (HEU). ...

  1. PRODUCTION OF URANIUM AND THORIUM COMPOUNDS

    DOE Patents [OSTI]

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

    1955-12-27

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

  2. Uranium Leasing Program: Program Summary | Department of Energy

    Energy Savers [EERE]

    lease tracts and offered to the domestic uranium industry through a competitive bid process. ... of DOE-managed lands for the exploration and production of uranium and vanadium ores. ...

  3. 1st Quarter 2016 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    2014,1242179,1095011,1468608,1085534,4891332 2015,1154408,789980,774541,624278,3343207 "P2016",626522,"NA","NA","NA","--" "E Estimated data. P Preliminary data. ...

  4. Domestic Uranium Production Report 2nd Quarter 2016

    Gasoline and Diesel Fuel Update (EIA)

    ... (Wyoming) 4. Ross CPP (Wyoming) 5. Smith Ranch-Highland Operation (Wyoming) 6. ... Resources, Inc. dba Cameco Resources Smith Ranch-Highland Operation Converse, Wyoming ...

  5. Accelerated Depletion: Assessing Its Impacts on Domestic Oil and Natural Gas Prices and Production

    Reports and Publications (EIA)

    2000-01-01

    Analysis of the potential impacts of accelerated depletion on domestic oil and natural gas prices and production.

  6. SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

    DOE Patents [OSTI]

    Spence, R.; Lister, M.W.

    1958-12-16

    Uranium and plutonium can be separated from neutron-lrradiated uranium by a process consisting of dissolvlng the lrradiated material in nitric acid, saturating the solution with a nitrate salt such as ammonium nitrate, rendering the solution substantially neutral with a base such as ammonia, adding a reducing agent such as hydroxylamine to change plutonium to the trivalent state, treating the solution with a substantially water immiscible organic solvent such as dibutoxy diethylether to selectively extract the uranium, maklng the residual aqueous solutlon acid with nitric acid, adding an oxidizing agent such as ammonlum bromate to oxidize the plutonium to the hexavalent state, and selectlvely extracting the plutonium by means of an immlscible solvent, such as dibutoxy dlethyletber.

  7. SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS

    DOE Patents [OSTI]

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

    1958-10-01

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

  8. IRON COATED URANIUM AND ITS PRODUCTION

    DOE Patents [OSTI]

    Gray, A.G.

    1960-03-15

    A method of applying a protective coating to a metallic uranium article is given. The method comprises etching the surface of the article with an etchant solution containlng chloride ions, such as a solution of phosphoric acid and hydrochloric acid, cleaning the etched surface, electroplating iron thereon from a ferrous ammonium sulfate electroplating bath, and soldering an aluminum sheath to the resultant iron layer.

  9. Table 5. Domestic Crude Oil Production, Projected vs. Actual

    U.S. Energy Information Administration (EIA) Indexed Site

    Domestic Crude Oil Production, Projected vs. Actual Projected (million barrels) 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 AEO 1994 2508 2373 2256 2161 2088 2022 1953 1891 1851 1825 1799 1781 1767 1759 1778 1789 1807 1862 AEO 1995 2402 2307 2205 2095 2037 1967 1953 1924 1916 1905 1894 1883 1887 1887 1920 1945 1967 AEO 1996 2387 2310 2248 2172 2113 2062 2011 1978 1953 1938 1916 1920 1927 1949 1971 1986 2000 2018 2055 AEO 1997 2362 2307

  10. Method for the production of uranium chloride salt

    DOE Patents [OSTI]

    Westphal, Brian R.; Mariani, Robert D.

    2013-07-02

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

  11. SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY ADSORPTION

    DOE Patents [OSTI]

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

    1958-01-01

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

  12. Fact# 904: December 21, 2015 Gross Domestic Product and Vehicle Travel:

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Both Increased during 2015 | Department of Energy 4: December 21, 2015 Gross Domestic Product and Vehicle Travel: Both Increased during 2015 Fact# 904: December 21, 2015 Gross Domestic Product and Vehicle Travel: Both Increased during 2015 SUBSCRIBE to the Fact of the Week The nation's highway vehicle miles of travel (VMT) and the U.S. gross domestic product (GDP) reflect strikingly similar patterns, indicating the strong relationship between the nation's economy and its travel. Beginning in

  13. Fact #904: December 21, 2015 Gross Domestic Product and Vehicle Travel:

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Both Increased during 2015 - Dataset | Department of Energy Fact #904: December 21, 2015 Gross Domestic Product and Vehicle Travel: Both Increased during 2015 - Dataset Fact #904: December 21, 2015 Gross Domestic Product and Vehicle Travel: Both Increased during 2015 - Dataset Excel file and dataset for Gross Domestic Product and Vehicle Travel: Both Increased during 2015 fotw#904_web_rev.xlsx (19.75 KB) More Documents & Publications Vehicle Technologies Office Spring 2016 Quarterly

  14. Fact #564: March 30, 2009 Transportation and the Gross Domestic Product,

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    2007 | Department of Energy 4: March 30, 2009 Transportation and the Gross Domestic Product, 2007 Fact #564: March 30, 2009 Transportation and the Gross Domestic Product, 2007 Transportation plays a major role in the U.S. economy. About 10% of the U.S. Gross Domestic Product (GDP) in 2007 is related to transportation. Housing, health care, and food are the only categories with greater shares of the GDP. GDP by Category, 2007 Graph showing the Gross Domestic Product (GDP) for various

  15. PROCESS FOR SEGREGATING URANIUM FROM PLUTONIUM AND FISSION-PRODUCT CONTAMINATION

    DOE Patents [OSTI]

    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.

  16. S. 2415: Title I may be cited as the Uranium Enrichment Act of 1990; Title II may be cited as the Uranium Security and Tailings Reclamation Act of 1989; and Title III may be cited as The Solar, Wind, Waste, and Geothermal Power Production Incentives Act of 1990, introduced in the Senate, One Hundred First Congress, Second Session, April 4, 1990

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    S. 2415 (which started out as a bill to encourage solar and geothermal power generation) now would amend the Atomic Energy Act of 1954 to redirect uranium enrichment enterprises to further the national interest, respond to competitive market forces, and to ensure the nation's common defense and security. It would establish a United States Enrichment Corporation for the following purposes: to acquire feed materials, enriched uranium, and enrichment facilities; to operate these facilities; to market enriched uranium for governmental purposes and qualified domestic and foreign persons; to conduct research into uranium enrichment; and to operate as a profitable, self-financing, reliable corporation and in a manner consistent with the health and safety of the public. The bill describes powers and duties of the corporation; the organization, finance, and management; decontamination and decommissioning. The second part of the bill would ensure an adequate supply of domestic uranium for defense and power production; provide assistance to the domestic uranium industry; and establish, facilitate, and expedite a comprehensive system for financing reclamation and remedial action at active uranium and thorium processing sites. The third part of the bill would remove the size limitations on power production facilities now part of the Public Utility Regulatory Policies Act of 1978. Solar, wind, waste, or geothermal power facilities would no longer have to be less than 80 MW to qualify as a small power production facility.

  17. uranium

    National Nuclear Security Administration (NNSA)

    to prepare surplus plutonium for disposition, and readiness to begin the Second Uranium Cycle, to start processing spent nuclear fuel.

    H Canyon is also being...

  18. Method for making a uranium chloride salt product

    DOE Patents [OSTI]

    Miller, William E.; Tomczuk, Zygmunt

    2004-10-05

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

  19. SEPARATION OF PLUTONIUM VALUES FROM URANIUM AND FISSION PRODUCT VALUES

    DOE Patents [OSTI]

    Maddock, A.G.; Booth, A.H.

    1960-09-13

    Separation of plutonium present in small amounts from neutron irradiated uranium by making use of the phenomenon of chemisorption is described. Plutonium in the tetravalent state is chemically absorbed on a fluoride in solid form. The steps for the separation comprise dissolving the irradiated uranium in nitric acid, oxidizing the plutonium in the resulting solution to the hexavalent state, adding to the solution a soluble calcium salt which by the common ion effect inhibits dissolution of the fluoride by the solution, passing the solution through a bed or column of subdivided calcium fluoride which has been sintered to about 8OO deg C to remove the chemisorbable fission products, reducing the plutonium in the solution thus obtained to the tetravalent state, and again passing the solution through a similar bed or column of calcium fluoride to selectively absorb the plutonium, which may then be recovered by treating the calcium fluoride with a solution of ammonium oxalate.

  20. New Technologies that Enhance Environmental Protection, Increase Domestic Production, Result from DOE-Supported Consortium

    Broader source: Energy.gov [DOE]

    New technologies that help small, independent oil and natural gas operators contribute to domestic energy production while improving environmental protection have resulted from U.S. Department of Energy support of the Stripper Well Consortium.

  1. India's Worsening Uranium Shortage

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2007-01-15

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

  2. Uranium production in Eastern Europe and its environmental impact: A literature survey

    SciTech Connect (OSTI)

    Norman, R.E.

    1993-04-01

    A survey of the unclassified literature was made to determine the location, technology, throughput, and environmental status of the uranium mines and mills that have historically made up uranium production capability in Eastern Europe. Included in that survey were the following countries: the former German Democratic Republic (GDR), now part of a reunited Germany, Czechoslovakia, Romania, Bulgaria, Hungary, and Poland. Until recently, uranium was being produced in five of these six countries (Poland stopped production 20 years ago). The production began directly after World War II in support of weapons production in the Soviet Union. Eastern Europe has produced about two-thirds of the total Soviet uranium inventory historically, or about 330,000 metric tonnes of uranium (NM) [730 million pounds of uranium (MlbU)l out of a total of about 490,000 MTU (1090 NlbU).

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

    DOE Patents [OSTI]

    Herrmann, Steven Douglas

    2014-05-27

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

  4. EIS-0360: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio

    Energy Savers [EERE]

    Site | Department of Energy 60: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio Site EIS-0360: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio Site Summary This site-specific EIS analyzes the construction, operation, maintenance, and decontamination and decommissioning of the proposed depleted uranium hexafluoride (DUF6) conversion facility at three alternative locations within the Portsmouth site; transportation of all cylinders (DUF6, enriched, and

  5. PROCESS FOR THE PRODUCTION OF AMMONIUM URANIUM FLUORIDE

    DOE Patents [OSTI]

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

    1953-08-25

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

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

    DOE Patents [OSTI]

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

    1957-09-17

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

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

    DOE Patents [OSTI]

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

    1959-08-18

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

  8. U.S. Domestic Oil Production Exceeds Imports for First Time in 18 Years |

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Department of Energy Domestic Oil Production Exceeds Imports for First Time in 18 Years U.S. Domestic Oil Production Exceeds Imports for First Time in 18 Years November 15, 2013 - 3:47pm Addthis Source: Energy Information Administration Short Term Energy Outlook. Chart by Daniel Wood. Allison Lantero Allison Lantero Digital Content Specialist, Office of Public Affairs In February 1995, The Brady Bunch Movie and Billy Madison were in movie theaters, "Creep" by TLC was at the top of

  9. Proceedings of Workshop on Uranium Production Environmental Restoration: An exchange between the United States and Germany

    SciTech Connect (OSTI)

    Not Available

    1993-12-31

    Scientists, engineers, elected officials, and industry regulators from the United, States and Germany met in Albuquerque, New Mexico, August 16--20, 1993, in the first joint international workshop to discuss uranium tailings remediation. Entitled ``Workshop on Uranium Production Environmental Restoration: An Exchange between the US and Germany,`` the meeting was hosted by the US Department of Energy`s (DOE) Uranium Mill Tailings Remedial Action (UMTRA) Project. The goal of the workshop was to further understanding and communication on the uranium tailings cleanup projects in the US and Germany. Many communities around the world are faced with an environmental legacy -- enormous quantities of hazardous and low-level radioactive materials from the production of uranium used for energy and nuclear weapons. In 1978, the US Congress passed the Uranium Mill Tailings Radiation Control Act. Title I of the law established a program to assess the tailings at inactive uranium processing sites and provide a means for joint federal and state funding of the cleanup efforts at sites where all or substantially all of the uranium was produced for sale to a federal agency. The UMTRA Project is responsible for the cleanup of 24 sites in 10 states. Germany is facing nearly identical uranium cleanup problems and has established a cleanup project. At the workshop, participants had an opportunity to interact with a broad cross section of the environmental restoration and waste disposal community, discuss common concerns and problems, and develop a broader understanding of the issues. Abstracts are catalogued individually for the data base.

  10. Process for Low Cost Domestic Production of LIB Cathode Materials

    SciTech Connect (OSTI)

    Thurston, Anthony

    2012-10-31

    The objective of the research was to determine the best low cost method for the large scale production of the Nickel-Cobalt-Manganese (NCM) layered cathode materials. The research and development focused on scaling up the licensed technology from Argonne National Laboratory in BASFs battery material pilot plant in Beachwood Ohio. Since BASF did not have experience with the large scale production of the NCM cathode materials there was a significant amount of development that was needed to support BASFs already existing research program. During the three year period BASF was able to develop and validate production processes for the NCM 111, 523 and 424 materials as well as begin development of the High Energy NCM. BASF also used this time period to provide free cathode material samples to numerous manufactures, OEMs and research companies in order to validate the ma-terials. The success of the project can be demonstrated by the construction of the production plant in Elyria Ohio and the successful operation of that facility. The benefit of the project to the public will begin to be apparent as soon as material from the production plant is being used in electric vehicles.

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

    DOE Patents [OSTI]

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

    1961-06-20

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

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

    SciTech Connect (OSTI)

    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 high 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.

  13. Draft Supplement Analysis for Location(s) to Dispose of Depleted Uranium Oxide Conversion Product Generated from DOE'S Inventory of Depleted Uranium Hexafluoride

    Office of Environmental Management (EM)

    DRAFT SUPPLEMENT ANALYSIS FOR LOCATION(S) TO DISPOSE OF DEPLETED URANIUM OXIDE CONVERSION PRODUCT GENERATED FROM DOE'S INVENTORY OF DEPLETED URANIUM HEXAFLUORIDE (DOE/EIS-0359-SA1 AND DOE/EIS-0360-SA1) March 2007 March 2007 i CONTENTS NOTATION........................................................................................................................... iv 1 INTRODUCTION AND BACKGROUND ................................................................. 1 1.1 Why DOE Has Prepared This

  14. The US uranium industry: Regulatory and policy impediments

    SciTech Connect (OSTI)

    Drennen, T.E.; Glicken, J.

    1995-06-01

    The Energy Policy Act of 1992 required the DOE to develop recommendations and implement government programs to assist the domestic uranium industry in increasing export opportunities. In 1993, as part of that effort, the Office of Nuclear Energy identified several key factors that could (or have) significantly impact(ed) export opportunities for domestic uranium. This report addresses one of these factors: regulatory and policy impediments to the flow of uranium products between the US and other countries. It speaks primarily to the uranium market for civil nuclear power. Changes in the world political and economic order have changed US national security requirements, and the US uranium industry has found itself without the protected market it once enjoyed. An unlevel playing field for US uranium producers has resulted from a combination of geology, history, and a general US political philosophy of nonintervention that precludes the type of industrial policy practiced in other uranium-exporting countries. The US has also been hampered in its efforts to support the domestic uranium-producing industry by its own commitment to free and open global markets and by international agreements such as GATT and NAFTA. Several US policies, including the imposition of NRC fees and licensing costs and Harbor Maintenance fees, directly harm the competitiveness of the domestic uranium industry. Finally, requirements under US law, such as those in the 1979 Nuclear Nonproliferation Act, place very strict limits on the use of US-origin uranium, limitations not imposed by other uranium-producing countries. Export promotion and coordination are two areas in which the US can help the domestic uranium industry without violating existing trade agreements or other legal or policy constraints.

  15. Economics of large-scale thorium oxide production: assessment of domestic resources

    SciTech Connect (OSTI)

    Young, J.K.; Bloomster, C.H.; Enderlin, W.I.; Morgenstern, M.H.; Ballinger, M.Y.; Drost, M.K.; Weakley, S.A.

    1980-02-01

    The supply curve illustrates that sufficient amounts of thorium exist supply a domestic thorium-reactor economy. Most likely costs of production range from $3 to $60/lb ThO/sub 2/. Near-term thorium oxide resources include the stockpiles in Ohio, Maryland, and Tennessee and the thorite deposits at Hall Mountain, Idaho. Costs are under $10/lb thorium oxide. Longer term economic deposits include Wet Mountain, Colorado; Lemhi Pass, Idaho; and Palmer, Michigan. Most likely costs are under $20/lb thorium oxide. Long-term deposits include Bald Mountain, Wyoming; Bear Lodge, Wyoming; and Conway, New Hampshire. Costs approximately equal or exceed $50/lb thorium oxide.

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

    SciTech Connect (OSTI)

    McGinnis, Brent

    2014-04-01

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

  17. Uranium industry annual 1996

    SciTech Connect (OSTI)

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs.

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

    SciTech Connect (OSTI)

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

    2000-02-29

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

  19. Background Fact Sheet Transfer of Depleted Uranium and Subsequent...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    The Department's National Nuclear Security Administration (NNSA) requires U.S.-origin unobligated domestically enriched, domestic-origin uranium to support continued tritium ...

  20. Rescuing a Treasure Uranium-233

    SciTech Connect (OSTI)

    Krichinsky, Alan M; Goldberg, Dr. Steven A.; Hutcheon, Dr. Ian D.

    2011-01-01

    Uranium-233 (233U) is a synthetic isotope of uranium formed under reactor conditions during neutron capture by natural thorium (232Th). At high purities, this synthetic isotope serves as a crucial reference for accurately quantifying and characterizing natural uranium isotopes for domestic and international safeguards. Separated 233U is stored in vaults at Oak Ridge National Laboratory. These materials represent a broad spectrum of 233U from the standpoint isotopic purity the purest being crucial for precise analyses in safeguarding uranium. All 233U at ORNL currently is scheduled to be down blended with depleted uranium beginning in 2015. Such down blending will permanently destroy the potential value of pure 233U samples as certified reference material for use in uranium analyses. Furthermore, no replacement 233U stocks are expected to be produced in the future due to a lack of operating production capability and the high cost of returning to operation this currently shut down capability. This paper will describe the efforts to rescue the purest of the 233U materials arguably national treasures from their destruction by down blending.

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

    DOE Patents [OSTI]

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

    1960-08-23

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

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

    DOE Patents [OSTI]

    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.

  3. METHOD OF SEPARATING FISSION PRODUCTS FROM FUSED BISMUTH-CONTAINING URANIUM

    DOE Patents [OSTI]

    Wiswall, R.H.

    1958-06-24

    A process is described for removing metal selectively from liquid metal compositions. The method effects separation of flssion product metals selectively from dilute solution in fused bismuth, which contains uraniunn in solution without removal of more than 1% of the uranium. The process comprises contacting the fused bismuth with a fused salt composition consisting of sodium, potassium and lithium chlorides, adding to fused bismuth and molten salt a quantity of bismuth chloride which is stoichiometrically required to convert the flssion product metals to be removed to their chlorides which are more stable in the fused salt than in the molten metal and are, therefore, preferentially taken up in the fused salt phase.

  4. Niowave Develops Production Route for Medical Radioisotopes with...

    Office of Science (SC) Website

    The lack of a domestic supply and current use of techniques requiring weapons grade uranium led Congress to pass the American Medical Isotope Production Act in 2013. The Act ...

  5. Table 2. Real Gross Domestic Product Growth Trends, Projected vs. Actual

    U.S. Energy Information Administration (EIA) Indexed Site

    Real Gross Domestic Product Growth Trends, Projected vs. Actual Projected Real GDP Growth Trend (cumulative average percent growth in projected real GDP from first year shown for each AEO) 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 AEO 1994 3.09 3.15 2.86 2.78 2.73 2.65 2.62 2.60 2.56 2.53 2.52 2.49 2.45 2.41 2.40 2.36 2.32 2.29 AEO 1995 3.66 2.77 2.53 2.71 2.67 2.61 2.55 2.48 2.46 2.45 2.45 2.43 2.39 2.35 2.31 2.27 2.24 AEO 1996 2.61

  6. Fact #828: July 7, 2014 Japanese Auto Manufacturers Increase Domestic Production for U.S. Sales

    Broader source: Energy.gov [DOE]

    In 1980, all Japanese-brand vehicles sold in the U.S. were imported. By 1990, just over one-third of Japanese-brand vehicles sold in the U.S. were produced domestically in North America which...

  7. DECONTAMINATION OF URANIUM

    DOE Patents [OSTI]

    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.

  8. OPEC and lower oil prices: Impacts on production capacity, export refining, domestic demand and trade balances

    SciTech Connect (OSTI)

    Fesharaki, F.; Fridley, D.; Isaak, D.; Totto, L.; Wilson, T.

    1988-12-01

    The East-West Center has received a research grant from the US Department of Energy's Office of Policy, Planning, and Analysis to study the impact of lower oil prices on OPEC production capacity, on export refineries, and petroleum trade. The project was later extended to include balance-of-payments scenarios and impacts on OPEC domestic demand. As the study progressed, a number of preliminary presentations were made at the US Department of Energy in order to receive feedback from DOE officials and to refine the focus of our analysis. During one of the presentations on June 4, 1987, the then Director of Division of Oil and Gas, John Stanley-Miller, advised us to focus our work on the Persian Gulf countries, since these countries were of special interest to the United States Government. Since then, our team has visited Iran, the United Arab Emirates, and Saudi Arabia and obtained detailed information from other countries. The political turmoil in the Gulf, the Iran/Iraq war, and the active US military presence have all worked to delay the final submission of our report. Even in countries where the United States has close ties, access to information has been difficult. In most countries, even mundane information on petroleum issues are treated as national secrets. As a result of these difficulties, we requested a one-year no cost extension to the grant and submitted an Interim Report in May 1988. As part of our grant extension request, we proposed to undertake additional tasks which appear in this report. 20 figs., 21 tabs.

  9. Method for converting uranium oxides to uranium metal

    DOE Patents [OSTI]

    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.

  10. Site evaluations for the uranium-atomic vapor laser isotope separation (U-AVLIS) production plant

    SciTech Connect (OSTI)

    Wolsko, T.; Absil, M.; Cirillo, R.; Folga, S.; Gillette, J.; Habegger, L.; Whitfield, R.

    1991-07-01

    This report describes a uranium-atomic vapor laser isotope separation (U-AVLIS) production plant siting study conducted during 1990 to identify alternative plant sites for examination in later environmental impact studies. A siting study methodology was developed in early 1990 and was implemented between June and December. This methodology had two parts. The first part -- a series of screening analyses that included exclusionary and other criteria -- was conducted to identify a reasonable number of candidates sites. This slate of candidate sites was then subjected to more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. To fully appreciate the siting study methodology, it is important to understand the U-AVLIS program and site requirements. 16 refs., 29 figs., 54 tabs.

  11. Uranium industry annual 1994

    SciTech Connect (OSTI)

    1995-07-05

    The Uranium Industry Annual 1994 (UIA 1994) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing during that survey year. The UIA 1994 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the 10-year period 1985 through 1994 as collected on the Form EIA-858, ``Uranium Industry Annual Survey.`` Data collected on the ``Uranium Industry Annual Survey`` (UIAS) provide a comprehensive statistical characterization of the industry`s activities for the survey year and also include some information about industry`s plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1994, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. A feature article, ``Comparison of Uranium Mill Tailings Reclamation in the United States and Canada,`` is included in the UIA 1994. Data on uranium raw materials activities including exploration activities and expenditures, EIA-estimated resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities, including purchases of uranium and enrichment services, and uranium inventories, enrichment feed deliveries (actual and projected), and unfilled market requirements are shown in Chapter 2.

  12. Uranium industry annual 1998

    SciTech Connect (OSTI)

    1999-04-22

    The Uranium Industry Annual 1998 (UIA 1998) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. It contains data for the period 1989 through 2008 as collected on the Form EIA-858, ``Uranium Industry Annual Survey.`` Data provides a comprehensive statistical characterization of the industry`s activities for the survey year and also include some information about industry`s plans and commitments for the near-term future. Data on uranium raw materials activities for 1989 through 1998, including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment, are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2008, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, and uranium inventories, are shown in Chapter 2. The methodology used in the 1998 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ``Uranium Industry Annual Survey`` is provided in Appendix C. The Form EIA-858 ``Uranium Industry Annual Survey`` is shown in Appendix D. For the readers convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix E along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 24 figs., 56 tabs.

  13. Evaluation of the Acceptability of Potential Depleted Uranium Hexafluoride Conversion Products at the Envirocare Disposal Site

    SciTech Connect (OSTI)

    Croff, A.G.

    2001-01-11

    The purpose of this report is to review and document the capability of potential products of depleted UF{sub 6} conversion to meet the current waste acceptance criteria and other regulatory requirements for disposal at the facility in Clive, Utah, owned by Envirocare of Utah, Inc. The investigation was conducted by identifying issues potentially related to disposal of depleted uranium (DU) products at Envirocare and conducting an initial analysis of them. Discussions were then held with representatives of Envirocare, the state of Utah (which is a NRC Agreement State and, thus, is the cognizant regulatory authority for Envirocare), and DOE Oak Ridge Operations. Provisional issue resolution was then established based on the analysis and discussions and documented in a draft report. The draft report was then reviewed by those providing information and revisions were made, which resulted in this document. Issues that were examined for resolution were (1) license receipt limits for U isotopes; (2) DU product classification as Class A waste; (3) use of non-DOE disposal sites for disposal of DOE material; (4) historical NRC views; (5) definition of chemical reactivity; (6) presence of mobile radionuclides; and (7) National Environmental Policy Act coverage of disposal. The conclusion of this analysis is that an amendment to the Envirocare license issued on October 5, 2000, has reduced the uncertainties regarding disposal of the DU product at Envirocare to the point that they are now comparable with uncertainties associated with the disposal of the DU product at the Nevada Test Site that were discussed in an earlier report.

  14. Statistical data of the uranium industry

    SciTech Connect (OSTI)

    1980-01-01

    This document is a compilation of historical facts and figures through 1979. These statistics are based primarily on information provided voluntarily by the uranium exploration, mining, and milling companies. The production, reserves, drilling, and production capability information has been reported in a manner which avoids disclosure of proprietary information. Only the totals for the $1.5 reserves are reported. Because of increased interest in higher cost resources for long range planning purposes, a section covering the distribution of $100 per pound reserves statistics has been newly included. A table of mill recovery ranges for the January 1, 1980 reserves has also been added to this year's edition. The section on domestic uranium production capability has been deleted this year but will be included next year. The January 1, 1980 potential resource estimates are unchanged from the January 1, 1979 estimates.

  15. SOLVENT EXTRACTION PROCESS FOR THE SEPARATION OF URANIUM AND THORIUM FROM PROTACTINIUM AND FISSION PRODUCTS

    DOE Patents [OSTI]

    Rainey, R.H.; Moore, J.G.

    1962-08-14

    A liquid-liquid extraction process was developed for recovering thorium and uranium values from a neutron irradiated thorium composition. They are separated from a solvent extraction system comprising a first end extraction stage for introducing an aqueous feed containing thorium and uranium into the system consisting of a plurality of intermediate extractiorr stages and a second end extractron stage for introducing an aqueous immiscible selective organic solvent for thorium and uranium in countercurrent contact therein with the aqueous feed. A nitrate iondeficient aqueous feed solution containing thorium and uranium was introduced into the first end extraction stage in countercurrent contact with the organic solvent entering the system from the second end extraction stage while intro ducing an aqueous solution of salting nitric acid into any one of the intermediate extraction stages of the system. The resultant thorium and uranium-laden organic solvent was removed at a point preceding the first end extraction stage of the system. (AEC)

  16. ESTABLISHMENT OF AN INDUSTRY-DRIVEN CONSORTIUM FOCUSED ON IMPROVING THE PRODUCTION PERFORMANCE OF DOMESTIC STRIPPER WELLS

    SciTech Connect (OSTI)

    Joel L. Morrison

    2004-05-17

    The Pennsylvania State University, under contract to the U.S. Department of Energy, National Energy Technology Laboratory will establish, promote, and manage a national industry-driven Stripper Well Consortium (SWC) that will be focused on improving the production performance of domestic petroleum and/or natural gas stripper wells. The consortium creates a partnership with the U.S. petroleum and natural gas industries and trade associations, state funding agencies, academia, and the National Energy Technology Laboratory. This report serves as the thirteenth quarterly technical progress report for the SWC. Key activities for this reporting period included: (1) hosting three fall technology transfer meetings in Wyoming, Texas, and Pennsylvania, (2) releasing the 2004 SWC request-for-proposal (RFP), and (3) initial planning of the SWC spring meeting in Golden Colorado for selecting the 2004 SWC projects. The Fall technology transfer meetings attracted 100+ attendees between the three workshops. The SWC membership which attended the Casper, Wyoming workshop was able to see several SWC-funded projects operating in the field at the Rocky Mountain Oilfield Testing Center. The SWC is nearing the end of its initial funding cycle. The Consortium has a solid membership foundation and a demonstrated ability to review and select projects that have relevancy to meet the needs of domestic stripper well operators.

  17. Effect of Co-solutes on the Products and Solubility of Uranium(VI) Precipitated with Phosphate

    SciTech Connect (OSTI)

    Mehta, Vrajesh; Maillot, Fabien; Wang, Zheming; Catalano, Jeffrey G.; Giammar, Daniel E.

    2014-01-22

    Uranyl phosphate solids are often found with uranium ores, and their low solubility makes them promising target phases for in situ remediation of uranium-contaminated subsurface environments. The products and solubility of uranium(VI) precipitated with phosphate can be affected by the pH, dissolved inorganic carbon (DIC) concentration, and co-solute composition (e.g. Na+/Ca2+) of the groundwater. Batch experiments were performed to study the effect of these parameters on the products and extent of uranium precipitation induced by phosphate addition. In the absence of co-solute cations, chernikovite [H3O(UO2)(PO4)•3H2O] precipitated despite uranyl orthophosphate [(UO2)3(PO4)2•4H2O] being thermodynamically more favorable under certain conditions. As determined using X-ray diffraction, electron microscopy, and laser induced fluorescence spectroscopy, the presence of Na+ or Ca2+ as a co-solute led to the precipitation of sodium autunite ([Na2(UO2)2(PO4)2] and autunite [Ca(UO2)2(PO4)2]), which are structurally similar to chernikovite. In the presence of sodium, the dissolved U(VI) concentrations were generally in agreement with equilibrium predictions of sodium autunite solubility. However, in the calcium-containing systems, the observed concentrations were below the predicted solubility of autunite, suggesting the possibility of uranium adsorption to or incorporation in a calcium phosphate precipitate in addition to the precipitation of autunite.

  18. Establishment of an Industry-Driven Consortium Focused on Improving the Production Performance of Domestic Stripper Wells

    SciTech Connect (OSTI)

    Joel L. Morrison; Sharon L. Elder

    2006-05-01

    The Pennsylvania State University, under contract to the U.S. Department of Energy (DOE), National Energy Technology Laboratory (NETL) established a national industry-driven Stripper Well Consortium (SWC) that is focused on improving the production performance of domestic petroleum and/or natural gas stripper wells. The consortium creates a partnership with the U.S. petroleum and natural gas industries and trade associations, state funding agencies, academia, and the National Energy Technology Laboratory. This report serves as the eighth quarterly technical progress report for the SWC. Key activities for this reporting period include: (1) Organize and host the 2006 Spring Meeting in State College, PA to review and select projects for SWC co-funding; (2) Participation in the 2006 PA CleanEnergy Expo Energy Theater to air the DVD on ''Independent Oil: Rediscovering American's Forgotten Wells''; (3) New member additions; (4) Improving communications; and (5) Planning of the fall technology meetings.

  19. ESTABLISHMENT OF AN INDUSTRY-DRIVEN CONSORTIUM FOCUSED ON IMPROVING THE PRODUCTION PERFORMANCE OF DOMESTIC STRIPPER WELLS

    SciTech Connect (OSTI)

    Joel L. Morrison

    2004-12-28

    The Pennsylvania State University, under contract to the U.S. Department of Energy, National Energy Technology Laboratory will establish, promote, and manage a national industry-driven Stripper Well Consortium (SWC) that will be focused on improving the production performance of domestic petroleum and/or natural gas stripper wells. The consortium creates a partnership with the U.S. petroleum and natural gas industries and trade associations, state funding agencies, academia, and the National Energy Technology Laboratory. This report serves as the first quarterly technical progress report for the SWC. Key activities for this reporting period include: (1) hosting the SWC spring proposal meeting in Golden Colorado, (2) planning of the upcoming SWC fall technology transfer meetings, and (3) recruiting the SWC base membership.

  20. ESTABLISHMENT OF AN INDUSTRY-DRIVEN CONSORTIUM FOCUSED ON IMPROVING THE PRODUCTION PERFORMANCE OF DOMESTIC STRIPPER WELLS

    SciTech Connect (OSTI)

    Joel L. Morrison

    2004-12-23

    The Pennsylvania State University, under contract to the U.S. Department of Energy, National Energy Technology Laboratory will establish, promote, and manage a national industry-driven Stripper Well Consortium (SWC) that will be focused on improving the production performance of domestic petroleum and/or natural gas stripper wells. The consortium creates a partnership with the U.S. petroleum and natural gas industries and trade associations, state funding agencies, academia, and the National Energy Technology Laboratory. This report serves as the fifteenth quarterly technical progress report for the SWC. Key activities for this reporting period include: (1) hosting the SWC spring proposal meeting in Golden Colorado, (2) planning of the upcoming SWC fall technology transfer meetings, and (3) recruiting the SWC base membership.

  1. Establishment of an Industry-Driven Consortium Focused on Improving the Production Performance of Domestic Stripper Wells

    SciTech Connect (OSTI)

    Joel Morrison; Sharon Elder

    2006-01-24

    The Pennsylvania State University, under contract to the U.S. Department of Energy, National Energy Technology Laboratory will establish, promote, and manage a national industry-driven Stripper Well Consortium (SWC) that will be focused on improving the production performance of domestic petroleum and/or natural gas stripper wells. The consortium creates a partnership with the U.S. petroleum and natural gas industries and trade associations, state funding agencies, academia, and the National Energy Technology Laboratory. This report serves as the sixth quarterly technical progress report for the SWC. Key activities for this reporting period included: (1) Organized and hosted two technology transfer meetings; (2) Collaborated with the Pennsylvania Oil and Gas Association (POGAM) to host a Natural Gas Outlook conference in Pittsburgh, PA; (3) Provided a SWC presentation at the Interstate Oil and Gas Compact Commission (IOGCC) meeting in Jackson Hole, WY; and (4) Completed and released a stripper well industry documentary entitled: ''Independent Oil: Rediscovering America's Forgotten Wells''.

  2. Uranium industry annual 1995

    SciTech Connect (OSTI)

    1996-05-01

    The Uranium Industry Annual 1995 (UIA 1995) provides current statistical data on the U.S. uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1995 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the period 1986 through 2005 as collected on the Form EIA-858, ``Uranium Industry Annual Survey``. Data collected on the ``Uranium Industry Annual Survey`` provide a comprehensive statistical characterization of the industry`s plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1995, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. Data on uranium raw materials activities for 1986 through 1995 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2005, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. The methodology used in the 1995 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ``Uranium Industry Annual Survey`` is provided in Appendix C. For the reader`s convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix D along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 14 figs., 56 tabs.

  3. Method for the recovery of uranium values from uranium tetrafluoride

    DOE Patents [OSTI]

    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.

  4. Method for the recovery of uranium values from uranium tetrafluoride

    DOE Patents [OSTI]

    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.

  5. Excess Uranium Management | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Excess Uranium Management Excess Uranium Management Request for Information - July 2016 On July 19, 2016, the Department issued a Request for Information on the effects of DOE transfers of excess uranium on domestic uranium mining, conversion, and enrichment industries. The Request for Information established an August 18, 2016 deadline for the submission of written comments. The Request for Information is available here. On August 1, 2016, the Department extended the comment period to September

  6. Uranium hexafluoride liquid thermal expansion, elusive eutectic with hydrogen fluoride, and very first production using chlorine trifluoride

    SciTech Connect (OSTI)

    Rutledge, G.P.

    1991-12-31

    Three unusual incidents and case histories involving uranium hexafluoride in the enrichment facilities of the USA in the late 1940`s and early 1950`s are presented. The history of the measurements of the thermal expansion of liquids containing fluorine atoms within the molecule is reviewed with special emphasis upon uranium hexafluoride. A comparison is made between fluorinated esters, fluorocarbons, and uranium hexafluoride. The quantitative relationship between the thermal expansion coefficient, a, of liquids and the critical temperature, T{sub c} is presented. Uranium hexafluoride has an a that is very high in a temperature range that is used by laboratory and production workers - much higher than any other liquid measured. This physical property of UF{sub 6} has resulted in accidents involving filling the UF{sub 6} containers too full and then heating with a resulting rupture of the container. Such an incident at a uranium gaseous diffusion plant is presented. Production workers seldom {open_quotes}see{close_quotes} uranium hexafluoride. The movement of UF{sub 6} from one container to another is usually trailed by weight, not sight. Even laboratory scientists seldom {open_quotes}see{close_quotes} solid or liquid UF{sub 6} and this can be a problem at times. This inability to {open_quotes}see{close_quotes} the UF{sub 6}-HF mixtures in the 61.2{degrees}C to 101{degrees}C temperature range caused a delay in the understanding of the phase diagram of UF{sub 6}-HF which has a liquid - liquid immiscible region that made the eutectic composition somewhat elusive. Transparent fluorothene tubes solved the problem both for the UF{sub 6}-HF phase diagram as well as the UF{sub 6}-HF-CIF{sub 3} phase diagram with a miscibility gap starting at 53{degrees}C. The historical background leading to the first use of CIF{sub 3} to produce UF{sub 6} in both the laboratory and plant at K-25 is presented.

  7. VANE Uranium One JV | Open Energy Information

    Open Energy Info (EERE)

    VANE Uranium One JV Jump to: navigation, search Name: VANE-Uranium One JV Place: London, England, United Kingdom Zip: EC4V 6DX Product: JV between VANE Minerals Plc & Uranium One....

  8. Projections of the impact of expansion of domestic heavy oil production on the U.S. refining industry from 1990 to 2010. Topical report

    SciTech Connect (OSTI)

    Olsen, D.K.; Ramzel, E.B.; Strycker, A.R.; Guariguata, G.; Salmen, F.G.

    1994-12-01

    This report is one of a series of publications assessing the feasibility of increasing domestic heavy oil (10{degrees} to 20{degrees} API gravity) production. This report provides a compendium of the United States refining industry and analyzes the industry by Petroleum Administration for Defense District (PADD) and by ten smaller refining areas. The refining capacity, oil source and oil quality are analyzed, and projections are made for the U.S. refining industry for the years 1990 to 2010. The study used publicly available data as background. A linear program model of the U.S. refining industry was constructed and validated using 1990 U.S. refinery performance. Projections of domestic oil production (decline) and import of crude oil (increases) were balanced to meet anticipated demand to establish a base case for years 1990 through 2010. The impact of additional domestic heavy oil production, (300 MB/D to 900 MB/D, originating in select areas of the U.S.) on the U.S. refining complex was evaluated. This heavy oil could reduce the import rate and the balance of payments by displacing some imported, principally Mid-east, medium crude. The construction cost for refining units to accommodate this additional domestic heavy oil production in both the low and high volume scenarios is about 7 billion dollars for bottoms conversion capacity (delayed coking) with about 50% of the cost attributed to compliance with the Clean Air Act Amendment of 1990.

  9. ESTABLISHMENT OF AN INDUSTRY-DRIVEN CONSORTIUM FOCUSED ON IMPROVING THE PRODUCTION PERFORMANCE OF DOMESTIC STRIPPER WELLS

    SciTech Connect (OSTI)

    Joel L. Morrison

    2002-08-27

    The Pennsylvania State University, under contract to the U.S. Department of Energy, National Energy Technology Laboratory will establish, promote, and manage a national industry-driven Stripper Well Consortium (SWC) that will be focused on improving the production performance of domestic petroleum and/or natural gas stripper wells. The consortium creates a partnership with the U.S. petroleum and natural gas industries and trade associations, state funding agencies, academia, and the National Energy Technology Laboratory. This report serves as the sixth quarterly technical progress report for the SWC. Key activities for this reporting period include: (1) release of 2002 SWC request-for-proposal, (2) organized and hosted the Spring SWC meeting in Columbus, Ohio for membership proposal presentations and review; (3) tentatively scheduled the 2002 fall technology transfer meeting sites, and (4) continued to recruit additional Consortium members. In addition, a literature search that focuses on the use of lasers, microwaves, and acoustics for potential stripper well applications continued.

  10. Report A-94 T" PRODUCTION OF DgO FOR USE IN THE FISSION OP URANIUM

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    v_^ ' / * Report A-94 T" PRODUCTION OF DgO FOR USE IN THE FISSION OP URANIUM Report by Herold G. Urey, Arlstld V. Grosse, and George Walden, June 23, 1941 LEGAL N O T I C E This report was prepared as an account of Government sponsored work. Neither the United States, nor the Commission, nor any person acting on behalf of the Commission: A . Makes any warranty or representation, express or implied, with respect to the a c - curacy, completeness, or usefulness of the information contained in

  11. ESTABLISHMENT OF AN INDUSTRY-DRIVEN CONSORTIUM FOCUSED ON IMPROVING THE PRODUCTION PERFORMANCE OF DOMESTIC STRIPPER WELLS

    SciTech Connect (OSTI)

    Joel L. Morrison

    2002-09-30

    The Pennsylvania State University, under contract to the U.S. Department of Energy (DOE), National Energy Technology Laboratory (NETL), has established a national industry-driven Stripper Well Consortium (SWC) that is focused on improving the production performance of domestic petroleum and/or natural gas stripper wells. The consortium creates a partnership with the U.S. petroleum and natural gas industries and trade associations, state funding agencies, academia, and the National Energy Technology Laboratory. This report serves as the second topical report. The SWC has grown and diversified its membership during its first 24 months of existence. The Consortium is now focused on building strategic alliances with additional industrial, state, and federal entities to expand further the SWC membership base and transfer technologies as they are developed. In addition, the Consortium has successfully worked to attract state support to co-fund SWC projects. Penn State has entered a co-funding arrangement with the New York State Energy Development Authority (NYSERDA) which has provided $200,000 over the last two years to co-fund stripper well production-orientated projects that have relevance to New York state producers. During this reporting period, the Executive Council approved co-funding for 14 projects that have a total project value of $2,116,897. Since its inception, the SWC has approved cofunding for 27 projects that have a total project value of $3,632,109.84. The SWC has provided $2,242,701 in co-funding for these projects and programmatically maintains a cost share of 39%.

  12. Production of small uranium dioxide microspheres for cermet nuclear fuel using the internal gelation process

    SciTech Connect (OSTI)

    Collins, Robert T; Collins, Jack Lee; Hunt, Rodney Dale; Ladd-Lively, Jennifer L; Patton, Kaara K; Hickman, Robert

    2014-01-01

    The U.S. National Aeronautics and Space Administration (NASA) is developing a uranium dioxide (UO2)/tungsten cermet fuel for potential use as the nuclear cryogenic propulsion stage (NCPS). The first generation NCPS is expected to be made from dense UO2 microspheres with diameters between 75 and 150 m. Previously, the internal gelation process and a hood-scale apparatus with a vibrating nozzle were used to form gel spheres, which became UO2 kernels with diameters between 350 and 850 m. For the NASA spheres, the vibrating nozzle was replaced with a custom designed, two-fluid nozzle to produce gel spheres in the desired smaller size range. This paper describes the operational methodology used to make 3 kg of uranium oxide microspheres.

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

    DOE Patents [OSTI]

    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.

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

    DOE Patents [OSTI]

    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.

  15. DECONTAMINATION OF URANIUM

    DOE Patents [OSTI]

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

    1962-05-15

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

  16. Method for fabricating .sup.99 Mo production targets using low enriched uranium, .sup.99 Mo production targets comprising low enriched uranium

    DOE Patents [OSTI]

    Wiencek, Thomas C.; Matos, James E.; Hofman, Gerard L.

    2000-12-12

    A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.

  17. Method for fabricating .sup.99 Mo production targets using low enriched uranium, .sup.99 Mo production targets comprising low enriched uranium

    DOE Patents [OSTI]

    Wiencek, Thomas C.; Matos, James E.; Hofman, Gerard L.

    1997-01-01

    A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.

  18. Method for fabricating {sup 99}Mo production targets using low enriched uranium, {sup 99}Mo production targets comprising low enriched uranium

    DOE Patents [OSTI]

    Wiencek, T.C.; Matos, J.E.; Hofman, G.L.

    1997-03-25

    A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate. 3 figs.

  19. Replacement Cost of Domestic Crude

    Energy Science and Technology Software Center (OSTI)

    1994-12-01

    The DEEPWATER model forecasts the replacement cost of domestic crude oil for 13 offshore regions in the lower 48 states. The replacement cost of domestic crude oil is the constant or levelized selling price that will recover the full expense of exploration, development, and productions with a reasonable return on capital.

  20. Management Controls over the Department of Energy's Uranium Leasing Program, OAS-M-08-05

    Broader source: Energy.gov [DOE]

    The Department of Energy's Uranium Leasing Program was established by the Atomic Energy Act of 1954 to develop a supply of domestic uranium to meet the nation's defense needs. Pursuant to the Act,...

  1. DOE Releases Excess Uranium Inventory Plan | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Excess Uranium Inventory Plan DOE Releases Excess Uranium Inventory Plan December 16, 2008 - 8:51am Addthis WASHINGTON, D.C. - The United States Department of Energy (DOE) today issued its Excess Uranium Inventory Management Plan (the Plan), which outlines the Department's strategy for the management and disposition of its excess uranium inventories. The Plan highlights DOE's ongoing efforts to enhance national security and promote a healthy domestic nuclear infrastructure through the efficient

  2. U.S. Domestic

    U.S. Energy Information Administration (EIA) Indexed Site

    Domestic and Foreign Coal Distribution by State of Origin ...Energy Information Administration | Annual Coal Distribution Report 2013 Domestic and ...

  3. Uranium immobilization and nuclear waste

    SciTech Connect (OSTI)

    Duffy, C.J.; Ogard, A.E.

    1982-02-01

    Considerable information useful in nuclear waste storage can be gained by studying the conditions of uranium ore deposit formation. Further information can be gained by comparing the chemistry of uranium to nuclear fission products and other radionuclides of concern to nuclear waste disposal. Redox state appears to be the most important variable in controlling uranium solubility, especially at near neutral pH, which is characteristic of most ground water. This is probably also true of neptunium, plutonium, and technetium. Further, redox conditions that immobilize uranium should immobilize these elements. The mechanisms that have produced uranium ore bodies in the Earth's crust are somewhat less clear. At the temperatures of hydrothermal uranium deposits, equilibrium models are probably adequate, aqueous uranium (VI) being reduced and precipitated by interaction with ferrous-iron-bearing oxides and silicates. In lower temperature roll-type uranium deposits, overall equilibrium may not have been achieved. The involvement of sulfate-reducing bacteria in ore-body formation has been postulated, but is uncertain. Reduced sulfur species do, however, appear to be involved in much of the low temperature uranium precipitation. Assessment of the possibility of uranium transport in natural ground water is complicated because the system is generally not in overall equilibrium. For this reason, Eh measurements are of limited value. If a ground water is to be capable of reducing uranium, it must contain ions capable of reducing uranium both thermodynamically and kinetically. At present, the best candidates are reduced sulfur species.

  4. URANIUM ALLOYS

    DOE Patents [OSTI]

    Colbeck, E.W.

    1959-12-29

    A uranium alloy is reported containing from 0.1 to 5 per cent by weight of molybdenum and from 0.1 to 5 per cent by weight of silicon, the balance being uranium.

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

    SciTech Connect (OSTI)

    Marmer, G.J.; Dunn, C.P.; Moeller, K.L.; Pfingston, J.M.; Policastro, A.J.; Yuen, C.R.; Cleland, J.H.

    1991-09-01

    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 part 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.

  6. Depleted uranium management alternatives

    SciTech Connect (OSTI)

    Hertzler, T.J.; Nishimoto, D.D.

    1994-08-01

    This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process.

  7. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

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

    1959-07-28

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

  8. URANIUM EXTRACTION

    DOE Patents [OSTI]

    Harrington, C.D.; Opie, J.V.

    1958-07-01

    The recovery of uranium values from uranium ore such as pitchblende is described. The ore is first dissolved in nitric acid, and a water soluble nitrate is added as a salting out agent. The resulting feed solution is then contacted with diethyl ether, whereby the bulk of the uranyl nitrate and a portion of the impurities are taken up by the ether. This acid ether extract is then separated from the aqueous raffinate, and contacted with water causing back extractioa of the uranyl nitrate and impurities into the water to form a crude liquor. After separation from the ether extract, this crude liquor is heated to about 118 deg C to obtain molten uranyl nitrate hexahydratc. After being slightly cooled the uranyl nitrate hexahydrate is contacted with acid free diethyl ether whereby the bulk of the uranyl nitrate is dissolved into the ethcr to form a neutral ether solution while most of the impurities remain in the aqueous waste. After separation from the aqueous waste, the resultant ether solution is washed with about l0% of its volume of water to free it of any dissolved impurities and is then contacted with at least one half its volume of water whereby the uranyl nitrate is extracted into the water to form an aqueous product solution.

  9. Electrolysis of uranium nitride containing fission product elements (Mo, Pd, Nd) in a molten LiCl-KCl eutectic

    SciTech Connect (OSTI)

    Satoh, Takumi; Iwai, Takashi; Arai, Yasuo

    2007-07-01

    The electrolysis of burnup-simulated uranium nitride, UN, containing representative solid fission product elements (Mo, Pd, Nd) was investigated in the molten LiCl-KCl eutectic salt with 0.54 wt% UCl{sub 3} from the view point of application of pyrochemical reprocessing to nitride fuel cycle. It was found from cyclic voltammetry and anodic polarization curve measurement that anodic dissolution of UN began at about -0.75 V vs. Ag/AgCl reference electrode in all samples. After the electrolysis at the constant anodic potential of -0.65 {approx} -0.60 V vs. Ag/AgCl, most of UN was dissolved into LiCl- KCl as UCl{sub 3} at the anode, and U was recovered in the liquid Cd cathode in all samples. Further, Nd was dissolved into LiCl-KCl as NdCl{sub 3}, while Mo and Pd were not dissolved but remained at the anode. (authors)

  10. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    5. Average price and quantity for uranium purchased by owners and operators of U.S. civilian nuclear power reactors by pricing mechanisms and delivery year, 2014-15 dollars per pound U3O8 equivalent; thousand pounds U3O8 equivalent Pricing mechanisms Domestic purchases1 Foreign purchases2 Total purchases 2014 2015 2014 2015 2014 2015 Contract-specified (fixed and base-escalated) pricing Weighted-average price 41.87 40.34 49.87 44.93 45.47 42.88 Quantity with reported price 15,711 13,862 12,815

  11. Evaluation of selected detector systems for products formed in the atmospheric hydrolysis of uranium hexafluoride

    SciTech Connect (OSTI)

    Bostick, W.D.; Bostick, D.T.

    1987-03-01

    Sensitive detection of UF/sub 6/ hydrolysis products, either by discontinuous sampling or by continuous or near real-time monitoring, is an important safety consideration for DOE contractors handling large quantities of UF/sub 6/. Automated continuous or rapid intermittent remote sensing of these reaction products can provide an alarm signal when a preselected threshold value has been exceeded (absolute response) or when a significant emission excursion has occurred (rate of change of response). This report evaluates the performance of selected devices for the detection of airborne materials formed in the release of liquid UF/sub 6/ (approx. =1.3 g) into an enclosed volume of 6 m/sup 3/; these experiments were initiated on October 23, 1986. The detection principles investigated are: photometric, gas detector tubes, and electrochemical sensor.

  12. ELECTROLYSIS OF THORIUM AND URANIUM

    DOE Patents [OSTI]

    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.

  13. URANIUM COMPOSITIONS

    DOE Patents [OSTI]

    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.

  14. Statistical data of the uranium industry

    SciTech Connect (OSTI)

    1981-01-01

    Data are presented on US uranium reserves, potential resources, exploration, mining, drilling, milling, and other activities of the uranium industry through 1980. The compendium reflects the basic programs of the Grand Junction Office. Statistics are based primarily on information provided by the uranium exploration, mining, and milling companies. Data on commercial U/sub 3/O/sub 8/ sales and purchases are included. Data on non-US uranium production and resources are presented in the appendix. (DMC)

  15. Uranium Processing Facility | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Uranium Processing Facility

  16. Microsoft Word - L15 01-22 Uranium Tranfers

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    To: Office of Nuclear Energy Department of Energy 1000 Independence Ave., SW Washington, DC 20585 From: Nan Swift Federal Affairs Manager National Taxpayers Union 108 N. Alfred Street Alexandria, VA 22314 Subject: Request for Information: Excess Uranium Management: Effects of DOE Transfers of Excess Uranium on Domestic Uranium Mining, Conversion, and Enrichment Industries To whom it may concern: On behalf of the members of the National Taxpayers Union (NTU), I write to express our concerns

  17. Uranium Mining, Conversion, and Enrichment Industries

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Analysis of Potential Impacts of Uranium Transfers on the Domestic Uranium Mining, Conversion, and Enrichment Industries May 1, 2015 ii EXECUTIVE SUMMARY: The Department of Energy ("Department" or "DOE") plans to transfer the equivalent of up to 2,100 metric tons ("MTU") of natural uranium per year (with a higher total for calendar year 2015, mainly because of transfers already executed or under way before today's determination). These transfers would include 1,600

  18. Uranium hexafluoride handling. Proceedings

    SciTech Connect (OSTI)

    Not Available

    1991-12-31

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF{sub 6} from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride. Selected papers were processed separately for inclusion in the Energy Science and Technology Database.

  19. Uranium industry annual 1997

    SciTech Connect (OSTI)

    1998-04-01

    This report provides statistical data on the U.S. uranium industry`s activities relating to uranium raw materials and uranium marketing.

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

    SciTech Connect (OSTI)

    Hall, Susan M.

    2013-09-15

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

  1. Uranium Processing Facility Site Readiness Subproject Completed...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ... Successfully Establishes Uranium Lease and Takeback Program to Support Critical Medical Isotope Production Apex Gold discussion fosters international cooperation in run-up to ...

  2. PREPARATION OF URANIUM TRIOXIDE

    DOE Patents [OSTI]

    Buckingham, J.S.

    1959-09-01

    The production of uranium trioxide from aqueous solutions of uranyl nitrate is discussed. The uranium trioxide is produced by adding sulfur or a sulfur-containing compound, such as thiourea, sulfamic acid, sulfuric acid, and ammonium sulfate, to the uranyl solution in an amount of about 0.5% by weight of the uranyl nitrate hexahydrate, evaporating the solution to dryness, and calcining the dry residue. The trioxide obtained by this method furnished a dioxide with a considerably higher reactivity with hydrogen fluoride than a trioxide prepared without the sulfur additive.

  3. The Next Generation Safeguards Initiative s High-Purity Uranium-233 Preservation Effort

    SciTech Connect (OSTI)

    Krichinsky, Alan M; Bostick, Debra A; Giaquinto, Joseph; Bayne, Charles; Goldberg, Dr. Steven A.; Humphrey, Dr. Marc; Hutcheon, Dr. Ian D.; Sobolev, Taissa

    2012-01-01

    High-purity 233U serves as a crucial reference material for accurately quantifying and characterizing uranium. The most accurate analytical results which can be obtained only with high-purity 233U certified reference material (CRM) are required when used to confirm compliance with international safeguards obligations and international nonproliferation agreements. The U.S. supply of 233U CRM is almost depleted, and existing domestic stocks of this synthetic isotope are scheduled to be down-blended for disposition with depleted uranium beginning in 2015. Down blending batches of high-purity 233U will permanently eliminate the value of this material as a CRM. Furthermore, no replacement 233U stocks are expected to be produced in the future due to a lack of operating production capability and the high cost of replacing such capability. Therefore, preserving select batches of high-purity 233U is of great value and will assist in retaining current analytical capabilities for uranium-bearing samples. Any organization placing a priority on accurate results of uranium analyses, or on the confirmation of trace uranium in environmental samples, has a vested interest in preserving this material. This paper describes the need for high-purity 233U, the consequences organizations and agencies face if this material is not preserved, and the progress and future plans for preserving select batches of the purest 233U materials from disposition. This work is supported by the Next Generation Safeguards Initiative, Office of Nonproliferation and International Security, National Nuclear Security Administration.

  4. JACKETING URANIUM

    DOE Patents [OSTI]

    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.

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

    SciTech Connect (OSTI)

    Marmer, G.J.; Dunn, C.P.; Filley, T.H.; Moeller, K.L.; Pfingston, J.M.; Policastro, A.J.; Cleland, J.H.

    1991-09-01

    Uranium enrichment in the United States has utilized a diffusion process to preferentially enrich the U-235 isotope in the uranium product. In the 1970s, the US Department of Energy (DOE) began investigating more efficient and cost-effective enrichment technologies. 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. Initial facility operation is anticipated for 1999. A programmatic document for use in screening DOE sites to locate a U-AVLIS production plant was developed and implemented in two parts. The first part consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. The final evaluation, which included sensitivity studies, identified the Oak Ridge Gaseous Diffusion Plant (ORGDP) site, the Paducah Gaseous Diffusion Plant (PGDP) site, and the Portsmouth Gaseous Diffusion Plant (PORTS) site as having significant advantages over the other sites considered. This environmental site description (ESD) provides a detailed description of the PORTS site and vicinity suitable for use in an environmental impact statement (EIS). This report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during site visits. 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.

  6. Domestic and Foreign Distribution

    U.S. Energy Information Administration (EIA) Indexed Site

    of U.S. Coal by State of Origin, 2008 Final May 2010 Domestic and Foreign Distribution of U.S. Coal by State of Origin, 2008 (Thousand Short Tons) State Region Domestic Foreign...

  7. U.S. Domestic

    Gasoline and Diesel Fuel Update (EIA)

    2 Domestic and Foreign Distribution of U.S. Coal by State of Origin, 2012 (thousand short tons) Coal Exports Coal Origin State and Region Domestic Distribution By Coal Mines By...

  8. U.S. Domestic

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    1 Domestic and foreign distribution of U.S. coal by State of origin, 2011 (thousand short tons) Coal Exports Coal Origin State and Region Domestic Distribution By Coal Mines By...

  9. Process for alloying uranium and niobium

    DOE Patents [OSTI]

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

    1991-01-01

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

  10. Removal of uranium from aqueous HF solutions

    DOE Patents [OSTI]

    Pulley, Howard; Seltzer, Steven F.

    1980-01-01

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

  11. Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center

    SciTech Connect (OSTI)

    Myers, Astasia

    2011-06-28

    The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.

  12. Future of the Department of Energy's uranium enrichment enterprise

    SciTech Connect (OSTI)

    Sewell, P.G.

    1991-11-01

    The national energy strategy (NES) developed at President Bush's direction provides a focus for the US Department of Energy (DOE) future policy and funding initiatives including those of the uranium enrichment enterprise. The NES identifies an important and continuing role for nuclear energy as part of a balanced array of energy sources for meeting US energy needs, especially the growing demand for electricity. For many years, growth in US electricity demand has exhibited a strong correlation with growth in gross national product. NEW projections indicate that the US will need between 190 and 275 GW of additional system capacity by 2010. In order to unable nuclear power to help meet this need, the NEW establishes basic objectives for nuclear power. These objectives are to have a first order of a new nuclear power plant by 1995 and to have such a plant operational by 2000. The expansion of nuclear power anticipated in the NEW affirms a continuing need for a strong domestic uranium enrichment services supply capability. In terms of the future outlook for uranium enrichment, the atomic vapor laser isotope separation (AVLIS) technology continues to hold great promise for commercial application. If AVLIS efforts are successful, significant financial benefits from the commercial use of AVLIS will be realized by customers and the AVLIS deployment entity by approximately the year 2000 and thereafter.

  13. Final Environmental assessment for the Uranium Lease Management Program

    SciTech Connect (OSTI)

    1995-07-01

    The US Department of Energy (DOE) has prepared a programmatic environmental assessment (EA) of the proposed action to continue leasing withdrawn lands and DOE-owned patented claims for the exploration and production of uranium and vanadium ores. The Domestic Uranium Program regulation, codified at Title 10, Part 760.1, of the US Code of Federal Regulations (CFR), gives DOE the flexibility to continue leasing these lands under the Uranium Lease Management Program (ULMP) if the agency determines that it is in its best interest to do so. A key element in determining what is in DOE`s ``best interest`` is the assessment of the environmental impacts that may be attributable to lease tract operations and associated activities. On the basis of the information and analyses presented in the EA for the ULMP, DOE has determined that the proposed action does not constitute a major Federal action significantly affecting the quality of the human environment, as defined in the National Environmental Policy Act (NEPA) of 1969 (42 United States Code 4321 et seq.), as amended.Therefore, preparation of an environmental impact statement is not required for the ULMP,and DOE is issuing this Finding, of No Significant Impact (FONSI).

  14. ELECTROCHEMICAL DECONTAMINATION AND RECOVERY OF URANIUM VALUES

    DOE Patents [OSTI]

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

    1958-05-13

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

  15. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

    This book presents the GAO's views on the Department of Energy's (DOE) program to develop a new uranium enrichment technology, the atomic vapor laser isotope separation process (AVLIS). Views are drawn from GAO's ongoing review of AVLIS, in which the technical, program, and market issues that need to be addressed before an AVLIS plant is built are examined.

  16. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Hyman, H.H.; Dreher, J.L.

    1959-07-01

    The recovery of uranium from the acidic aqueous metal waste solutions resulting from the bismuth phosphate carrier precipitation of plutonium from solutions of neutron irradiated uranium is described. The waste solutions consist of phosphoric acid, sulfuric acid, and uranium as a uranyl salt, together with salts of the fission products normally associated with neutron irradiated uranium. Generally, the process of the invention involves the partial neutralization of the waste solution with sodium hydroxide, followed by conversion of the solution to a pH 11 by mixing therewith sufficient sodium carbonate. The resultant carbonate-complexed waste is contacted with a titanated silica gel and the adsorbent separated from the aqueous medium. The aqueous solution is then mixed with sufficient acetic acid to bring the pH of the aqueous medium to between 4 and 5, whereby sodium uranyl acetate is precipitated. The precipitate is dissolved in nitric acid and the resulting solution preferably provided with salting out agents. Uranyl nitrate is recovered from the solution by extraction with an ether such as diethyl ether.

  17. METHOD OF PRODUCING URANIUM

    DOE Patents [OSTI]

    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.

  18. Beneficial Uses of Depleted Uranium

    SciTech Connect (OSTI)

    Brown, C.; Croff, A.G.; Haire, M. J.

    1997-08-01

    Naturally occurring uranium contains 0.71 wt% {sup 235}U. In order for the uranium to be useful in most fission reactors, it must be enriched the concentration of the fissile isotope {sup 235}U must be increased. Depleted uranium (DU) is a co-product of the processing of natural uranium to produce enriched uranium, and DU has a {sup 235}U concentration of less than 0.71 wt%. In the United States, essentially all of the DU inventory is in the chemical form of uranium hexafluoride (UF{sub 6}) and is stored in large cylinders above ground. If this co-product material were to be declared surplus, converted to a stable oxide form, and disposed, the costs are estimated to be several billion dollars. Only small amounts of DU have at this time been beneficially reused. The U.S. Department of Energy (DOE) has begun the Beneficial Uses of DU Project to identify large-scale uses of DU and encourage its reuse for the primary purpose of potentially reducing the cost and expediting the disposition of the DU inventory. This paper discusses the inventory of DU and its rate of increase; DU disposition options; beneficial use options; a preliminary cost analysis; and major technical, institutional, and regulatory issues to be resolved.

  19. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-08-01

    This paper reports that in 1990 the Department of Energy began a two-year project to illustrate the technical and economic feasibility of a new uranium enrichment technology-the atomic vapor laser isotope separation (AVLIS) process. GAO believes that completing the AVLIS demonstration project will provide valuable information about the technical viability and cost of building an AVLIS plant and will keep future plant construction options open. However, Congress should be aware that DOE still needs to adequately demonstrate AVLIS with full-scale equipment and develop convincing cost projects. Program activities, such as the plant-licensing process, that must be completed before a plant is built, could take many years. Further, an updated and expanded uranium enrichment analysis will be needed before any decision is made about building an AVLIS plant. GAO, which has long supported legislation that would restructure DOE's uranium enrichment program as a government corporation, encourages DOE's goal of transferring AVLIS to the corporation. This could reduce the government's financial risk and help ensure that the decision to build an AVLIS plant is based on commercial concerns. DOE, however, has no alternative plans should the government corporation not be formed. Further, by curtailing a planned public access program, which would have given private firms an opportunity to learn about the technology during the demonstration project, DOE may limit its ability to transfer AVLIS to the private sector.

  20. Electrochemical method of producing eutectic uranium alloy and apparatus

    DOE Patents [OSTI]

    Horton, James A.; Hayden, H. Wayne

    1995-01-01

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

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

    SciTech Connect (OSTI)

    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 series 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.

  2. COPPER COATED URANIUM ARTICLE

    DOE Patents [OSTI]

    Gray, A.G.

    1958-10-01

    Various techniques and methods for obtaining coppercoated uranium are given. Specifically disclosed are a group of complex uranium coatings having successive layers of nickel, copper, lead, and tin.

  3. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    2. Maximum anticipated uranium market requirements of owners and operators of U.S. ... Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual ...

  4. PROCESS FOR PRODUCING URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Fowler, R.D.

    1957-10-22

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

  5. Uranium Mining and Milling near Rifle, Colorado | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Uranium Mining and Milling near Rifle, Colorado Uranium Mining and Milling near Rifle, Colorado April 19, 2016 - 4:42pm Addthis What does this project do? Goal 1. Protect Human Health and the Environment The small town of Rifle, Colorado, has an interesting history related to uranium and vanadium production. A mineral found near Rifle, called roscolite, contains both vanadium and uranium but was originally mined and milled for its vanadium content. Union Carbide Corporation began milling the ore

  6. EIS-0359: Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky

    Energy Savers [EERE]

    Site | Department of Energy 59: Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky Site EIS-0359: Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky Site Summary This site-specific EIS considers the construction, operation, maintenance, and decontamination and decommissioning of the proposed depleted uranium hexafluoride (DUF6) conversion facility at three locations within the Paducah site; transportation of depleted uranium conversion products and waste

  7. Continuous reduction of uranium tetrafluoride

    SciTech Connect (OSTI)

    DeMint, A.L.; Maxey, A.W.

    1993-10-21

    Operation of a pilot-scale system for continuous metallothermic reduction of uranium tetrafluoride (UF{sub 4} or green salt) has been initiated. This activity is in support of the development of a cost- effective process to produce uranium-iron (U-Fe) alloy feed for the Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) program. To date, five runs have been made to reduce green salt (UF{sub 4}) with magnesium. During this quarter, three runs were made to perfect the feeding system, examine feed rates, and determine the need for a crust breaker/stirrer. No material was drawn off in any of the runs; both product metal and by-product salt were allowed to accumulate in the reactor.

  8. DOE - Office of Legacy Management -- Falls City Uranium Ore Stockpile - TX

    Office of Legacy Management (LM)

    04A Falls City Uranium Ore Stockpile - TX 04A FUSRAP Considered Sites Site: Falls City Uranium Ore Stockpile (TX.04A ) Designated Name: Alternate Name: Location: Evaluation Year: Site Operations: Site Disposition: Radioactive Materials Handled: Primary Radioactive Materials Handled: Radiological Survey(s): Site Status: The history of domestic uranium procurement under U.S. Atomic Energy Commission (AEC) contracts identifies a number of ore buying stations (sampling and storage sites) that

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

    SciTech Connect (OSTI)

    Ferrada, J.J.

    2000-04-03

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

  10. Reducing emissions from uranium dissolving

    SciTech Connect (OSTI)

    Griffith, W.L.; Compere, A.L.; Huxtable, W.P.; Googin, J.M.

    1992-10-01

    This study was designed to assess the feasibility of decreasing NO[sub x] emissions from the current uranium alloy scrap tray dissolving facility. In the current process, uranium scrap is dissolved in boiling nitric acid in shallow stainless-steel trays. As scrap dissolves, more metal and more nitric acid are added to the tray by operating personnel. Safe geometry is assured by keeping liquid level at or below 5 cm, the depth of a safe infinite slab. The accountability batch control system provides additional protection against criticality. Both uranium and uranium alloys are dissolved. Nitric acid is recovered from the vapors for reuse. Metal nitrates are sent to uranium recovery. Brown NO[sub x] fumes evolved during dissolving have occasionally resulted in a visible plume from the trays. The fuming is most noticeable during startup and after addition of fresh acid to a tray. Present environmental regulations are expected to require control of brown NO[sub x] emissions. A detailed review of the literature, indicated the feasibility of slightly altering process chemistry to favor the production of NO[sub 2] which can be scrubbed and recycled as nitric acid. Methods for controlling the process to manage offgas product distribution and to minimize chemical reaction hazards were also considered.

  11. Reducing emissions from uranium dissolving

    SciTech Connect (OSTI)

    Griffith, W.L.; Compere, A.L.; Huxtable, W.P.; Googin, J.M.

    1992-10-01

    This study was designed to assess the feasibility of decreasing NO{sub x} emissions from the current uranium alloy scrap tray dissolving facility. In the current process, uranium scrap is dissolved in boiling nitric acid in shallow stainless-steel trays. As scrap dissolves, more metal and more nitric acid are added to the tray by operating personnel. Safe geometry is assured by keeping liquid level at or below 5 cm, the depth of a safe infinite slab. The accountability batch control system provides additional protection against criticality. Both uranium and uranium alloys are dissolved. Nitric acid is recovered from the vapors for reuse. Metal nitrates are sent to uranium recovery. Brown NO{sub x} fumes evolved during dissolving have occasionally resulted in a visible plume from the trays. The fuming is most noticeable during startup and after addition of fresh acid to a tray. Present environmental regulations are expected to require control of brown NO{sub x} emissions. A detailed review of the literature, indicated the feasibility of slightly altering process chemistry to favor the production of NO{sub 2} which can be scrubbed and recycled as nitric acid. Methods for controlling the process to manage offgas product distribution and to minimize chemical reaction hazards were also considered.

  12. Potentiometric determination of uranium in organic extracts

    SciTech Connect (OSTI)

    Bodnar, L.Z.

    1980-05-01

    The potentimetric determination of uranium in organic extracts was studied. A mixture of 30% TBP, (tributylphosphate), in carbon tetrachloride was used, with the NBL (New Brunswick Laboratory) titrimetric procedure. Results include a comparative analysis performed on organic extracts of fissium alloys vs those performed on aqueous samples of the same alloys which had been treated to remove interfering elements. Also comparative analyses were performed on sample solutions from a typical scrap recovery operation common in the uranium processing industry. A limited number of residue type materials, calciner products, and presscakes were subjected to analysis by organic extraction. The uranium extraction was not hindered by 30% TBP/CCl/sub 4/. To fully demonstrate the capabilities of the extraction technique and its compatibility with the NBL potentiometric uranium determination, a series of uranium standards was subjected to uranium extraction with 30% TBP/CCl/sub 4/. The uranium was then stripped out of the organic phase with 40 mL of H/sub 3/PO/sub 4/, 15 mL of H/sub 2/0, and 1 mL of 1M FeSO/sub 4/ solution. The uranium was then determined in the aqueous phosphoric phase by the regular NBL potentiometric method, omitting only the addition of another 40 mL of H/sub 3/PO/sub 4/. Uranium determinations ranging from approximately 20 to 150 mg of U were successfully made with the same accuracy and precision normally achieved. 8 tables. (DP)

  13. Preserving Ultra-Pure Uranium-233

    SciTech Connect (OSTI)

    Krichinsky, Alan M; Goldberg, Dr. Steven A.; Hutcheon, Dr. Ian D.

    2011-10-01

    Uranium-233 ({sup 233}U) is a synthetic isotope of uranium formed under reactor conditions during neutron capture by natural thorium ({sup 232}Th). At high purities, this synthetic isotope serves as a crucial reference material for accurately quantifying and characterizing uranium-bearing materials assays and isotopic distributions for domestic and international nuclear safeguards. Separated, high purity {sup 233}U is stored in vaults at Oak Ridge National Laboratory (ORNL). These materials represent a broad spectrum of {sup 233}U from the standpoint of isotopic purity - the purest being crucial for precise analyses in safeguarding uranium. All {sup 233}U at ORNL is currently scheduled to be disposed of by down-blending with depleted uranium beginning in 2015. This will reduce safety concerns and security costs associated with storage. Down-blending this material will permanently destroy its potential value as a certified reference material for use in uranium analyses. Furthermore, no credible options exist for replacing {sup 233}U due to the lack of operating production capability and the high cost of restarting currently shut down capabilities. A study was commissioned to determine the need for preserving high-purity {sup 233}U. This study looked at the current supply and the historical and continuing domestic need for this crucial isotope. It examined the gap in supplies and uses to meet domestic needs and extrapolated them in the context of international safeguards and security activities - superimposed on the recognition that existing supplies are being depleted while candidate replacement material is being prepared for disposal. This study found that the total worldwide need by this projection is at least 850 g of certified {sup 233}U reference material over the next 50 years. This amount also includes a strategic reserve. To meet this need, 18 individual items totaling 959 g of {sup 233}U were identified as candidates for establishing a lasting supply of

  14. URANIUM DECONTAMINATION

    DOE Patents [OSTI]

    Buckingham, J.S.; Carroll, J.L.

    1959-12-22

    A process is described for reducing the extractability of ruthenium, zirconium, and niobium values into hexone contained in an aqueous nitric acid uranium-containing solution. The solution is made acid-deficient, heated to between 55 and 70 deg C, and at that temperature a water-soluble inorganic thiosulfate is added. By this, a precipitate is formed which carries the bulk of the ruthenium, and the remainder of the ruthenium as well as the zirconium and niobium are converted to a hexone-nonextractable form. The rutheniumcontaining precipitate can either be removed from the solu tion or it can be dissolved as a hexone-non-extractable compound by the addition of sodium dichromate prior to hexone extraction.

  15. 2015 Annual Site Inspection and Monitoring Report for Uranium...

    Office of Legacy Management (LM)

    ... which had been a production well for the Bluewater mill, had a uranium concentration above the drinking water standard but below limits considered safe for livestock consumption. ...

  16. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    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.

  17. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    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.

  18. LITERATURE REVIEW ON THE SORPTION OF PLUTONIUM, URANIUM, NEPTUNIUM, AMERICIUM AND TECHNETIUM TO CORROSION PRODUCTS ON WASTE TANK LINERS

    SciTech Connect (OSTI)

    Li, D.; Kaplan, D.

    2012-02-29

    The Savannah River Site (SRS) has conducted performance assessment (PA) calculations to determine the risk associated with closing liquid waste tanks. The PA estimates the risk associated with a number of scenarios, making various assumptions. Throughout all of these scenarios, it is assumed that the carbon-steel tank liners holding the liquid waste do not sorb the radionuclides. Tank liners have been shown to form corrosion products, such as Fe-oxyhydroxides (Wiersma and Subramanian 2002). Many corrosion products, including Fe-oxyhydroxides, at the high pH values of tank effluent, take on a very strong negative charge. Given that many radionuclides may have net positive charges, either as free ions or complexed species, it is expected that many radionuclides will sorb to corrosion products associated with tank liners. The objective of this report was to conduct a literature review to investigate whether Pu, U, Np, Am and Tc would sorb to corrosion products on tank liners after they were filled with reducing grout (cementitious material containing slag to promote reducing conditions). The approach was to evaluate radionuclides sorption literature with iron oxyhydroxide phases, such as hematite ({alpha}-Fe{sub 2}O{sub 3}), magnetite (Fe{sub 3}O{sub 4}), goethite ({alpha}-FeOOH) and ferrihydrite (Fe{sub 2}O{sub 3} {center_dot} 0.5H{sub 2}O). The primary interest was the sorption behavior under tank closure conditions where the tanks will be filled with reducing cementitious materials. Because there were no laboratory studies conducted using site specific experimental conditions, (e.g., high pH and HLW tank aqueous and solid phase chemical conditions), it was necessary to extend the literature review to lower pH studies and noncementitious conditions. Consequently, this report relied on existing lower pH trends, existing geochemical modeling, and experimental spectroscopic evidence conducted at lower pH levels. The scope did not include evaluating the appropriateness

  19. Resource Assessment for Hydrogen Production: Hydrogen Production Potential from Fossil and Renewable Energy Resources

    SciTech Connect (OSTI)

    Melaina, M.; Penev, M.; Heimiller, D.

    2013-09-01

    This study examines the energy resources required to produce 4-10 million metric tonnes of domestic, low-carbon hydrogen in order to fuel approximately 20-50 million fuel cell electric vehicles. These projected energy resource requirements are compared to current consumption levels, projected 2040 business as usual consumptions levels, and projected 2040 consumption levels within a carbonconstrained future for the following energy resources: coal (assuming carbon capture and storage), natural gas, nuclear (uranium), biomass, wind (on- and offshore), and solar (photovoltaics and concentrating solar power). The analysis framework builds upon previous analysis results estimating hydrogen production potentials and drawing comparisons with economy-wide resource production projections

  20. Scrap uranium recycling via electron beam melting

    SciTech Connect (OSTI)

    McKoon, R.

    1993-11-01

    A program is underway at the Lawrence Livermore National Laboratory (LLNL) to recycle scrap uranium metal. Currently, much of the material from forging and machining processes is considered radioactive waste and is disposed of by oxidation and encapsulation at significant cost. In the recycling process, uranium and uranium alloys in various forms will be processed by electron beam melting and continuously cast into ingots meeting applicable specifications for virgin material. Existing vacuum processing facilities at LLNL are in compliance with all current federal and state environmental, safety and health regulations for the electron beam melting and vaporization of uranium metal. One of these facilities has been retrofitted with an auxiliary electron beam gun system, water-cooled hearth, crucible and ingot puller to create an electron beam melt furnace. In this furnace, basic process R&D on uranium recycling will be performed with the goal of eventual transfer of this technology to a production facility.

  1. Process for electroslag refining of uranium and uranium alloys

    DOE Patents [OSTI]

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

    1975-07-22

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

  2. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    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.

  3. Method of recovering uranium hexafluoride

    DOE Patents [OSTI]

    Schuman, S.

    1975-12-01

    A method of recovering uranium hexafluoride from gaseous mixtures which comprises adsorbing said uranium hexafluoride on activated carbon is described.

  4. NICKEL COATED URANIUM ARTICLE

    DOE Patents [OSTI]

    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.

  5. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    5. Shipments of uranium feed by owners and operators of U.S. civilian nuclear power ... Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual ...

  6. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    Inventories of uranium by owner as of end of year, 2011-15 thousand pounds U3O8 equivalent Inventories at the end of the year Owner of uranium inventory 2011 2012 2013 2014 P2015 ...

  7. Uranium Marketing Annual Report

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Uranium sellers to owners and operators of U.S. civilian nuclear power reactors, 2013-15 2013 2014 2015 American Fuel Resources, LLC Advance Uranium Asset Management Ltd. AREVA ...

  8. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    0. Contracted purchases of uranium from suppliers by owners and operators of U.S. civilian ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  9. Uranium Marketing Annual Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    a. Foreign purchases, foreign sales, and uranium inventories owned by U.S. suppliers and ... Foreign sales U.S. supplier owned uranium inventories Owners and operators of U.S. ...

  10. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors by year, 2011-15 thousand pounds U3O8 equivalent Origin of uranium 2011 2012 2013 2014 P2015 ...

  11. METHOD FOR PURIFYING URANIUM

    DOE Patents [OSTI]

    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.

  12. PROCESS OF PURIFYING URANIUM

    DOE Patents [OSTI]

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

    1958-12-23

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

  13. PREPARATION OF URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

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

    1959-10-01

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

  14. Electrolytic process for preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A.

    1990-01-01

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

  15. Statistical data of the uranium industry

    SciTech Connect (OSTI)

    1983-01-01

    This report is a compendium of information relating to US uranium reserves and potential resources and to exploration, mining, milling, and other activities of the uranium industry through 1982. The statistics are based primarily on data provided voluntarily by the uranium exploration, mining and milling companies. The compendium has been published annually since 1968 and reflects the basic programs of the Grand Junction Area Office of the US Department of Energy. Statistical data obtained from surveys conducted by the Energy Information Administration are included in Section IX. The production, reserves, and drilling data are reported in a manner which avoids disclosure of proprietary information.

  16. Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the Government of the United States and the Government of the Russian Federation has on the

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the Government of the United States of America and the Government of the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Operation of the Gaseous Diffusion Plant 2008 Information Date: December 31, 2008 1 Introduction The Agreement Between the Government of the United States of America and the Government of the Russian Federation

  17. Final Uranium Leasing Program Programmatic Environmental Impact...

    Energy Savers [EERE]

    Uranium Leasing Program Programmatic Environmental Impact Statement (PEIS) Final Uranium Leasing Program Programmatic Environmental Impact Statement (PEIS) Uranium Leasing ...

  18. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    Major U.S. Uranium Reserves

  19. The ultimate disposition of depleted uranium

    SciTech Connect (OSTI)

    Lemons, T.R.

    1991-12-31

    Depleted uranium (DU) is produced as a by-product of the uranium enrichment process. Over 340,000 MTU of DU in the form of UF{sub 6} have been accumulated at the US government gaseous diffusion plants and the stockpile continues to grow. An overview of issues and objectives associated with the inventory management and the ultimate disposition of this material is presented.

  20. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Yeager, J.H.

    1958-08-12

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

  1. URANIUM SEPARATION PROCESS

    DOE Patents [OSTI]

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

    1959-07-14

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

  2. PROCESS OF RECOVERING URANIUM

    DOE Patents [OSTI]

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

    1958-10-01

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

  3. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    SciTech Connect (OSTI)

    1995-07-05

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

  4. Reducing Emissions from Uranium Dissolving

    SciTech Connect (OSTI)

    Griffith, W.L.

    1992-01-01

    This study was designed to assess the feasibility of decreasing NO{sub x} emissions from the current uranium alloy scrap tray dissolving facility. In the current process, uranium scrap is dissolved in boiling nitric acid in shallow stainless-steel trays. As scrap dissolves, more metal and more nitric acid are added to the tray by operating personnel. Safe geometry is assured by keeping liquid level at or below 5 cm, the depth of a safe infinite slab. The accountability batch control system provides additional protection against criticality. The trays are steam coil heated. The process has operated satisfactorily, with few difficulties, for decades. Both uranium and uranium alloys are dissolved. Nitric acid is recovered from the vapors for reuse. Metal nitrates are sent to uranium recovery. Brown NO{sub x} fumes evolved during dissolving have occasionally resulted in a visible plume from the trays. The fuming is most noticeable during startup and after addition of fresh acid to a tray. Present environmental regulations are expected to require control of brown NO{sub x} emissions. Because NO{sub x} is hazardous, fumes should be suppressed whenever the electric blower system is inoperable. Because the tray dissolving process has worked well for decades, as much of the current capital equipment and operating procedures as possible were preserved. A detailed review of the literature, indicated the feasibility of slightly altering process chemistry to favor the production of NO{sub 2}, which can be scrubbed and recycled as nitric acid. Methods for controlling the process to manage offgas product distribution and to minimize chemical reaction hazards were also considered.

  5. Why Hydrogen? Hydrogen from Diverse Domestic Resources

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    from Diverse Domestic Resources Hydrogen from Diverse Domestic Resources Distributed Generation Transportation HIGH EFFICIENCY HIGH EFFICIENCY & RELIABILITY & RELIABILITY ZERONEAR...

  6. PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

    DOE Patents [OSTI]

    Moore, R.H.

    1962-10-01

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

  7. Fact #564: March 30, 2009 Transportation and the Gross Domestic...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Housing, health care, and food are the only categories with greater shares of the GDP. GDP ... Gross Domestic Product, 2007 Housing 24.3% Health Care 17.4% Food 11.6% ...

  8. :- : DRILLING URANIUM BILLETS ON A

    Office of Legacy Management (LM)

    'Xxy";^ ...... ' '. .- -- Metals, Ceramics, and Materials. : . - ,.. ; - . _ : , , ' z . , -, .- . >. ; . .. :- : DRILLING URANIUM BILLETS ON A .-... r .. .. i ' LEBLOND-CARLSTEDT RAPID BORER 4 r . _.i'- ' ...... ' -'".. :-'' ,' :... : , '.- ' ;BY R.' J. ' ANSEN .AEC RESEARCH AND DEVELOPMENT REPORT PERSONAL PROPERTY OF J. F. Schlltz .:- DECLASSIFIED - PER AUTHORITY OF (DAlE) (NhTI L (DATE)UE) FEED MATERIALS PRODUCTION CENTER NATIONAL LFE A COMPANY OF OHIO 26 1 3967 3035406 NLCO -

  9. About the Uranium Mine Team | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Uranium Mine Team About the Uranium Mine Team Text coming

  10. Construction of a Li Ion Battery (LIB) Cathode Production Plant...

    Broader source: Energy.gov (indexed) [DOE]

    Process for Low Cost Domestic Production of LIB Cathode Materials Process for Low Cost Domestic Production of LIB Cathode Materials Construction of a Li Ion Battery (LIB) Cathode ...

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

    SciTech Connect (OSTI)

    Rogers, Robin

    2013-12-21

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

  12. Preparation of uranium compounds

    DOE Patents [OSTI]

    Kiplinger, Jaqueline L; Montreal, Marisa J; Thomson, Robert K; Cantat, Thibault; Travia, Nicholas E

    2013-02-19

    UI.sub.3(1,4-dioxane).sub.1.5 and UI.sub.4(1,4-dioxane).sub.2, were synthesized in high yield by reacting turnings of elemental uranium with iodine dissolved in 1,4-dioxane under mild conditions. These molecular compounds of uranium are thermally stable and excellent precursor materials for synthesizing other molecular compounds of uranium including alkoxide, amide, organometallic, and halide compounds.

  13. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    S2. Uranium feed deliveries, enrichment services, and uranium loaded by owners and operators of U.S. civilian nuclear power reactors, 1994-2015 million pounds U3O8 equivalent million separative work units (SWU) Year Feed deliveries by owners and operators of U.S. civilian nuclear power reactors Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors U.S.-origin enrichment services purchased Foreign-origin enrichment services purchased Total purchased enrichment services

  14. Uranium Dispersion & Dosimetry Model.

    Energy Science and Technology Software Center (OSTI)

    2002-03-22

    The Uranium Dispersion and Dosimetry (UDAD) program provides estimates of potential radiation exposure to individuals and to the general population in the vicinity of a uranium processing facility such as a uranium mine or mill. Only transport through the air is considered. Exposure results from inhalation, external irradiation from airborne and ground-deposited activity, and ingestion of foodstuffs. Individual dose commitments, population dose commitments, and environmental dose commitments are computed. The program was developed for applicationmore » to uranium mining and milling; however, it may be applied to dispersion of any other pollutant.« less

  15. Uranium Purchases Report

    Reports and Publications (EIA)

    1996-01-01

    Final issue. This report details natural and enriched uranium purchases as reported by owners and operators of commercial nuclear power plants. 1996 represents the most recent publication year.

  16. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Kaufman, D.

    1958-04-15

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

  17. COATING URANIUM FROM CARBONYLS

    DOE Patents [OSTI]

    Gurinsky, D.H.; Storrs, S.S.

    1959-07-14

    Methods are described for making adherent corrosion resistant coatings on uranium metal. According to the invention, the uranium metal is heated in the presence of an organometallic compound such as the carbonyls of nickel, molybdenum, chromium, niobium, and tungsten at a temperature sufficient to decompose the metal carbonyl and dry plate the resultant free metal on the surface of the uranium metal body. The metal coated body is then further heated at a higher temperature to thermally diffuse the coating metal within the uranium bcdy.

  18. highly enriched uranium

    National Nuclear Security Administration (NNSA)

    and radioisotope supply capabilities of MURR and Nordion with General Atomics' selective gas extraction technology-which allows their low-enriched uranium (LEU) targets to remain...

  19. METHOD OF ROLLING URANIUM

    DOE Patents [OSTI]

    Smith, C.S.

    1959-08-01

    A method is described for rolling uranium metal at relatively low temperatures and under non-oxidizing conditions. The method involves the steps of heating the uranium to 200 deg C in an oil bath, withdrawing the uranium and permitting the oil to drain so that only a thin protective coating remains and rolling the oil coated uranium at a temperature of 200 deg C to give about a 15% reduction in thickness at each pass. The operation may be repeated to accomplish about a 90% reduction without edge cracking, checking or any appreciable increase in brittleness.

  20. Fermentation and Hydrogen Metabolism Affect Uranium Reduction by Clostridia

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    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 bymore » clostridia 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

  1. U.S.Uranium Reserves

    Gasoline and Diesel Fuel Update (EIA)

    Uranium Reserves Data for: 2003 Release Date: June 2004 Next Release: Not determined Uranium Reserves Estimates The Energy Information Administration (EIA) has reported the...

  2. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    1 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  3. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Minimum ...

  4. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Origin of ...

  5. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    7 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  6. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    3 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual ...

  7. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    9 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  8. PROCESS FOR MAKING URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Rosen, R.

    1959-07-14

    A process is described for producing uranium hexafluoride by reacting uranium hexachloride with hydrogen fluoride at a temperature below about 150 deg C, under anhydrous conditions.

  9. 2015 Uranium Market Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 2015 Uranium ... received in 2015 Weighted-average price Number of purchase contracts for ...

  10. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Number of purchasers Quantity with reported price ...

  11. URANIUM LEACHING AND RECOVERY PROCESS

    DOE Patents [OSTI]

    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.

  12. Uranium accountancy in Atomic Vapor Laser Isotope Separation

    SciTech Connect (OSTI)

    Carver, R.D.

    1986-01-01

    The AVLIS program pioneers the large scale industrial application of lasers to produce low cost enriched uranium fuel for light water reactors. In the process developed at Lawrence Livermore National Laboratory, normal uranium is vaporized by an electron beam, and a precisely tuned laser beam selectively photo-ionizes the uranium-235 isotopes. These ions are moved in an electromagnetic field to be condensed on the product collector. All other uranium isotopes remain uncharged and pass through the collector section to condense as tails. Tracking the three types of uranium through the process presents special problems in accountancy. After demonstration runs, the uranium on the collector was analyzed for isotopic content by Battelle Pacific Northwest Laboratory. Their results were checked at LLNL by analysis of parallel samples. The differences in isotopic composition as reported by the two laboratories were not significant.

  13. PROCESS OF PREPARING A FLUORIDE OF TETRAVLENT URANIUM

    DOE Patents [OSTI]

    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.

  14. Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy (DOE) owns and manages an inventory of depleted uranium (DU), natural uranium (NU), and low-enriched uranium (LEU) that is currently stored in large cylinders as...

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

    DOE Patents [OSTI]

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

    1982-06-29

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

  16. Synthesis of Uranium Trichloride for the Pyrometallurgical Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

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

    2011-11-01

    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 and conclusions from several experiments are presented along with a set of optimized operating conditions for the synthesis of uranium trichloride.

  17. URANIUM-SERIES CONSTRAINTS ON RADIONUCLIDE TRANSPORT AND GROUNDWATER FLOW AT NOPAL I URANIUM DEPOSIT, SIERRA PENA BLANCA, MEXICO

    SciTech Connect (OSTI)

    S. J. Goldstein, S. Luo, T. L. Ku, and M. T. Murrell

    2006-04-01

    Uranium-series data for groundwater samples from the vicinity of the Nopal I uranium ore deposit are used to place constraints on radionuclide transport and hydrologic processes at this site, and also, by analogy, at Yucca Mountain. Decreasing uranium concentrations for wells drilled in 2003 suggest that groundwater flow rates are low (< 10 m/yr). Field tests, well productivity, and uranium isotopic constraints also suggest that groundwater flow and mixing is limited at this site. The uranium isotopic systematics for water collected in the mine adit are consistent with longer rock-water interaction times and higher uranium dissolution rates at the front of the adit where the deposit is located. Short-lived nuclide data for groundwater wells are used to calculate retardation factors that are on the order of 1,000 for radium and 10,000 to 10,000,000 for lead and polonium. Radium has enhanced mobility in adit water and fractures near the deposit.

  18. Uranium-Loaded Water Treatment Resins: 'Equivalent Feed' at NRC and Agreement State-Licensed Uranium Recovery Facilities - 12094

    SciTech Connect (OSTI)

    Camper, Larry W.; Michalak, Paul; Cohen, Stephen; Carter, Ted

    2012-07-01

    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 and 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)

  19. URANIUM SEPARATION PROCESS

    DOE Patents [OSTI]

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

    1959-03-10

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

  20. Uranium dioxide electrolysis

    SciTech Connect (OSTI)

    Willit, James L.; Ackerman, John P.; Williamson, Mark A.

    2009-12-29

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

  1. METHOD FOR PURIFYING URANIUM

    DOE Patents [OSTI]

    Kennedy, J.W.; Segre, E.G.

    1958-08-26

    A method is presented for obtaining a compound of uranium in an extremely pure state and in such a condition that it can be used in determinations of the isotopic composition of uranium. Uranium deposited in calutron receivers is removed therefrom by washing with cold nitric acid and the resulting solution, coataining uranium and trace amounts of various impurities, such as Fe, Ag, Zn, Pb, and Ni, is then subjected to various analytical manipulations to obtain an impurity-free uranium containing solution. This solution is then evaporated on a platinum disk and the residue is ignited converting it to U2/sub 3//sub 8/. The platinum disk having such a thin film of pure U/sub 2/O/sub 8/ is suitable for use with isotopic determination techaiques.

  2. URANIUM PRECIPITATION PROCESS

    DOE Patents [OSTI]

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

    1957-12-01

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

  3. U.S. Energy Information Administration (EIA) - Ap

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Nuclear & Uranium view all Status of U.S. Nuclear Outages Daily statistics on nuclear capacity by plant. Domestic Uranium Production Report - Quarterly Information on uranium ...

  4. Domestic Health Studies and Activities | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Domestic Health Studies and Activities Domestic Health Studies and Activities Purpose The Atomic Energy Act of 1957 - Section 8(a) requires research and development activities relating to the protection of health during research and production activities. The requirement is fulfilled by conducting and supporting health studies and other research activities to determine if DOE workers and people living in communities near DOE sites are adversely affected by exposures to hazardous materials from

  5. Effect of Increased Natural Gas Exports on Domestic Energy Markets

    Reports and Publications (EIA)

    2012-01-01

    This report responds to an August 2011 request from the Department of Energy's Office of Fossil Energy (DOE\\/FE) for an analysis of "the impact of increased domestic natural gas demand, as exports." Appendix A provides a copy of the DOE\\/FE request letter. Specifically, DOE\\/FE asked the U.S. Energy Information Administration (EIA) to assess how specified scenarios of increased natural gas exports could affect domestic energy markets, focusing on consumption, production, and prices.

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

    SciTech Connect (OSTI)

    Francis, Stephen

    2013-07-01

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

  7. Environmental development plan for special nuclear materials production

    SciTech Connect (OSTI)

    Not Available

    1980-07-01

    The scope of this Environmental Development Plan (EDP) follows the Special Nuclear Materials (SNM) Production category of the Department of Energy budget. It includes the process steps and facilities necessary for the production of plutonium and tritium for Government needs and the production of some other radioactive materials that will be used for heat and radiation sources by domestic and international customers. Most of the SNM production is now carried out at the Savannah River Plant, but plutonium is still produced at Hanford. This EDP does not address the mining, milling, or enrichment of the uranium, but it does consider the reactor fuel (and target) fabrication facilities. The production reactors and the spent fuel processing plants and their effluents are discussed here, but the defense wastes from them are treated in a separate EDP. The scope does not include transportation, decontamination and decommissioning, safeguards and security, or use of the SNM products.

  8. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    SciTech Connect (OSTI)

    A K Wertsching

    2012-09-01

    As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to

  9. 2016 Billion-Ton Report: Advancing Domestic Resources for a Thriving

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Bioeconomy | Department of Energy Report: Advancing Domestic Resources for a Thriving Bioeconomy 2016 Billion-Ton Report: Advancing Domestic Resources for a Thriving Bioeconomy 2016 Billion-Ton Report: Advancing Domestic Resources for a Thriving Bioeconomy, Volume I Download the full interactive report to view visualizations of potential energy crop production, agricultural residues, forestry production and other scenarios on the BioenergyKDF. 2016_billion_ton_report.pdf (29.08 MB) More

  10. Electrochemical method of producing eutectic uranium alloy and apparatus

    DOE Patents [OSTI]

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

    1995-01-10

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

  11. PROCESS OF RECOVERING URANIUM

    DOE Patents [OSTI]

    Kilner, S.B.

    1959-12-29

    A method is presented for separating and recovering uranium from a complex mixure of impurities. The uranium is dissolved to produce an aqueous acidic solution including various impurities. In accordance with one method, with the uranium in the uranyl state, hydrogen cyanide is introduced into the solution to complex the impurities. Subsequently, ammonia is added to the solution to precipitate the uraniunn as ammonium diuranate away from the impurities in the solution. Alternatively, the uranium is precipitated by adding an alkaline metal hydroxide. In accordance with the second method, the uranium is reduced to the uranous state in the solution. The reduced solution is then treated with solid alkali metal cyanide sufficient to render the solution about 0.1 to 1.0 N in cyanide ions whereat cyanide complex ions of the metal impurities are produced and the uranium is simultaneously precipituted as uranous hydroxide. Alternatively, hydrogen cyanide may be added to the reduced solution and the uranium precipitated subsequently by adding ammonium hydroxide or an alkali metal hydroxide. Other refinements of the method are also disclosed.

  12. Advanced uranium enrichment technologies. Hearing before the Subcommittee on Energy Research and Production of the Committee on Science and Technology, US House of Representatives, Ninety-Sixth Congress, first session, September 22, 1979

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    This hearing was to learn about projected requirements for enriched uranium. The gas centrifuge work at Oak Ridge, Tennessee, and Portsmouth, Ohio, needed assessing. Laser isotope separation technique needed to be reviewed. Three technologies currently being emphasized in the Department of Energy's Advanced Isotope Separation (AIS) program were discussed; these included the Molecular Laser Isotope Separation (MLIS), Livermore's process called Atomic Vapor Laser Isotope Separation (AVLIS), and Plasma Separation Process (PSP). The status of each process was given. The present DOE AIS program calls for a process selection at the end of FY 1981, development module operation starting in the mid-1980's, pilot plant operations through the late 1980's and early 1990's, and a first production plant in the mid-1990's. (DP)

  13. Survey of lands held for uranium exploration, development, and production in fourteen western states in the six month period ending June 30, 1979

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    The statistics set forth for the period covered by this report are based on data gathered from records available to the public. These data were derived from public county records of mining claim locations, from the public reports of state and Federal land offices, from commercial reporting services, and from annual reports to stockholders of land companies. Accordingly, if any fee land has been acquired in a private transaction which has not been entered into a public record or report, that land will not be accounted for in this report. The figures for the acreage controlled at the beginning of the calendar year are those that were published for that date in the publication entitled Statistical Data of the Uranium Industry GJO-100(78).

  14. A Bill to amend the Internal Revenue Code of 1986 to provide incentives for domestic oil and natural gas exploration and production, and for other purposes. Introduced in the House of Representatives, One Hundred Third Congress, First Session, February 22, 1993

    SciTech Connect (OSTI)

    Not Available

    1993-01-01

    This Act may be cited as the [open quotes]Energy Independence, Infrastructure, and Investment Act of 1993[close quotes]. The purpose of this Bill is to amend the Internal Revenue Code of 1986 to provide incentives for domestic oil and natural gas exploration and production, and for other purposes. Title I of this Bill is Energy Independence Incentives. Title II is Infrastructure Incentives. Title III is Investment Incentives.

  15. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    4. Deliveries of uranium feed for enrichment by owners and operators of U.S. civilian nuclear power reactors by origin country and delivery year, 2013-15 thousand pounds U3O8 ...

  16. Uranium Marketing Annual Report -

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    9. Contracted purchases of uranium by owners and operators of U.S. civilian nuclear power reactors, signed in 2015, by delivery year, 2016-25 thousand pounds U3O8 equivalent Year ...

  17. PURIFICATION OF URANIUM FUELS

    DOE Patents [OSTI]

    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.

  18. ANODIC TREATMENT OF URANIUM

    DOE Patents [OSTI]

    Kolodney, M.

    1959-02-01

    A method is presented for effecting eloctrolytic dissolution of a metallic uranium article at a uniform rate. The uranium is made the anode in an aqueous phosphoric acid solution containing nitrate ions furnished by either ammonium nitrate, lithium nitrate, sodium nitrate, or potassium nitrate. A stainless steel cathode is employed and electrolysls carried out at a current density of about 0.1 to 1 ampere per square inch.

  19. URANIUM EXTRACTION PROCESS

    DOE Patents [OSTI]

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

    1958-12-16

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

  20. PROCESS FOR THE RECOVERY OF URANIUM FROM PHOSPHATIC ORE

    DOE Patents [OSTI]

    Long, R.L.

    1959-04-14

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

  1. Environmental control technology for mining and milling low-grade uranium resources

    SciTech Connect (OSTI)

    Weakley, S.A.; Blahnik, D.E.; Long, L.W.; Bloomster, C.H.

    1981-04-01

    This study examined the type and level of wastes that would be generated in the mining and milling of U/sub 3/O/sub 8/ from four potential domestic sources of uranium. The estimated costs of the technology to control these wastes to different degrees of stringency are presented.

  2. PROCESSING OF NEUTRON-IRRADIATED URANIUM

    DOE Patents [OSTI]

    Hopkins, H.H. Jr.

    1960-09-01

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

  3. Occurrence of Metastudtite (Uranium Peroxide Dihydrate) at a FUSRAP Site

    SciTech Connect (OSTI)

    Young, C.M.; Nelson, K.A.; Stevens, G.T.; Grassi, V.J.

    2006-07-01

    Uranium concentrations in groundwater in a localized area of a site exceed the USEPA Maximum Contaminant Level (MCL) by a factor of one thousand. Although the groundwater seepage velocity ranges up to 0.7 meters per day (m/day), data indicate that the uranium is not migrating in groundwater. We believe that the uranium is not mobile because of local geochemical conditions and the unstable nature of the uranium compound present at the site; uranium peroxide dihydrate (metastudtite). Metastudtite [UO{sub 4}.2(H{sub 2}O) or (U(O{sub 2})|O|(OH){sub 2}).3H{sub 2}O] has been identified at other sites as an alteration product in casks of spent nuclear fuel, but neither enriched nor depleted uranium were present at this site. Metastudtite was first identified as a natural mineral in 1983, although documented occurrences in the environment are uncommon. The U.S. Army Corps of Engineers (USACE) is conducting a remedial investigation at the DuPont Chambers Works in Deep water New Jersey under the Formerly Utilized Sites Remedial Action Program (FUSRAP) to evaluate radioactive contamination resulting from historical activities conducted in support of Manhattan Engineering District operations. From 1942 to 1947, Chambers Works converted uranium oxides to uranium tetrafluoride and uranium metal. More than half of the production at this facility resulted from the recovery process, where uranium-bearing dross and scrap were reacted with hydrogen peroxide to produce uranium peroxide dihydrate. The 280-hectare Chambers Works has produced some 600 products, including petrochemicals, aromatics, fluoro-chemicals, polymers, and elastomers. Contaminants resulting from these processes, including separate-phase petrochemicals, have also been detected within the boundaries of the FUSRAP investigation. USACE initiated remedial investigation field activities in 2002. The radionuclides of concern are natural uranium (U{sub nat}) and its short-lived progeny. Areas of impacted soil generally

  4. Assessment of Preferred Depleted Uranium Disposal Forms

    SciTech Connect (OSTI)

    Croff, A.G.; Hightower, J.R.; Lee, D.W.; Michaels, G.E.; Ranek, N.L.; Trabalka, J.R.

    2000-06-01

    The Department of Energy (DOE) is in the process of converting about 700,000 metric tons (MT) of depleted uranium hexafluoride (DUF6) containing 475,000 MT of depleted uranium (DU) to a stable form more suitable for long-term storage or disposal. Potential conversion forms include the tetrafluoride (DUF4), oxide (DUO2 or DU3O8), or metal. If worthwhile beneficial uses cannot be found for the DU product form, it will be sent to an appropriate site for disposal. The DU products are considered to be low-level waste (LLW) under both DOE orders and Nuclear Regulatory Commission (NRC) regulations. The objective of this study was to assess the acceptability of the potential DU conversion products at potential LLW disposal sites to provide a basis for DOE decisions on the preferred DU product form and a path forward that will ensure reliable and efficient disposal.

  5. Nuclear Fuel Facts: Uranium | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Facts: Uranium Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Uranium is a silvery-white metallic chemical element in the periodic table, with atomic number 92. It is assigned the chemical symbol U. A uranium atom has 92 protons and 92 electrons, of which 6 are valence electrons. Uranium has the highest atomic weight (19 kg m) of all naturally occurring elements. Uranium occurs naturally in low concentrations in soil, rock and water, and is commercially extracted from uranium-bearing

  6. Uranium Processing Facility Team Signs Partnering Agreement ...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Processing Facility ... Uranium Processing Facility Team Signs Partnering Agreement ... Nuclear Security, LLC; John Eschenberg, Uranium Processing Facility Project Office; Brian ...

  7. Survey of lands held for uranium exploration, development, and production in fourteen western states in the six-month period ending June 30, 1980

    SciTech Connect (OSTI)

    Not Available

    1981-01-01

    The statistics set forth for the period covered in this report are based on data gathered from records available to the public. The county records of mining claim locations, reports of state and federal land offices, and commercial reporting services furnish the data for this report. Accordingly, if any fee land has been acquired in a private transaction not entered into a public record or report, that land transaction will not be accounted for in this report. Manpower is not available to survey, acquire, and evaluate data from each available source in each reporting period. Therefore, in any given report, the figures quoted for one or more land categories in a given state may be identical to the figures shown in earlier reports even though some changes probably have occurred. Such changes will be shown on subsequent reports. The figures used for acreage controlled at the beginning of the calendar year are those published for that date in Statistical Data of the Uranium Industry GJ0-100 published and distributed by the Grand Junction Office of the Department of Energy.

  8. Survey of lands held for uranium exploration, development, and production in fourteen western states for the six-month period ending December 31, 1981

    SciTech Connect (OSTI)

    Not Available

    1982-04-01

    The statistics set forth for the period covered in this report are based on data gathered from records available to the public. The county records of mining claim locations, reports of state and federal land offices, and commercial reporting services furnish the data for this report. Accordingly, if any fee land has been acquired in a private transaction not entered into a public record or report, that, land transaction will not be accounted for in this report. Manpower is not available to survey, acquire, and evaluate data from each available source in each reporting period. Therefore, in any given report, the figures quoted for one or more land categories in a given state may be identical to figures shown in earlier reports even though some changes probably have occurred. Such changes will be shown on subsequent reports. The figures used for acreage controlled at the beginning of the calendar year are those published for that date in Statistical Data of the Uranium Industry GJO-100(81) published and distributed by the Grand Junction Area Office of the Department of Energy.

  9. Survey of lands held for uranium exploration, development, and production in fourteen western states in the six-month period ending December 31, 1979

    SciTech Connect (OSTI)

    Not Available

    1980-04-01

    The statistics set forth for the period covered in this report are based on data gathered from records available to the public. The county records of mining claim locations, reports of state and federal land offices, and commercial reporting services furnish the data for this report. Accordingly, if any fee land has been acquired in a private transaction not entered into a public record or report, that land transaction will not be accounted for in this report. Manpower is not available to survey, acquire, and evaluate data from each available source in each reporting period. Therefore, in any given report, the figures quoted for one or more land categories in a given state may be identical to the figures shown in earlier reports even though some changes probably have occurred. Such changes will be shown on subsequent reports. The figures used for acreage controlled at the beginning of the calendar year are those published for that date in Statistical Data of the Uranium Industry GJO-100 (79) published and distributed by the Grand Junction Office of the Department of Energy.

  10. Survey of lands held for uranium exploration, development, and production in fourteen western states for the six-month period ending June 30, 1981

    SciTech Connect (OSTI)

    Not Available

    1981-11-01

    The statistics set forth for the period covered in this report are based on data gathered from records available to the public. The county records of mining claim locations, reports of state and federal land offices, and commercial reporting services furnish the data for this report. Accordingly, if any fee land has been acquired in a private transaction not entered into a public record or report, that land transaction will not be accounted for in this report. Manpower is not available to survey, acquire, and evaluate data from each available source in each reporting period. Therefore, in any given report, the figures quoted for one or more land categories in a given state may be identical to the figures shown in earlier reports even though some changes probably have occurred. Such changes will be shown on subsequent reports. The figures used for acreage controlled at the beginning of the calender year are those published for that date in Statistical Data of the Uranium Industry GJO-100(81) published and distributed by the Grand Junction Area Office of the Department of Energy.

  11. Influence of uranium hydride oxidation on uranium metal behaviour

    SciTech Connect (OSTI)

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

    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, if 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)

  12. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    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.

  13. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    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.

  14. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

    Pruett, David J. (Knoxville, TN); McTaggart, Donald R. (Knoxville, TN)

    1984-01-01

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

  15. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

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

    1983-08-31

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

  16. WA_98_001_REYNOLDS_METALS_COMPANY_Waiver_of_Domestic_and_For.pdf |

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Department of Energy 8_001_REYNOLDS_METALS_COMPANY_Waiver_of_Domestic_and_For.pdf WA_98_001_REYNOLDS_METALS_COMPANY_Waiver_of_Domestic_and_For.pdf (1.1 MB) More Documents & Publications Advance Patent Waiver W(A)1998-014 WA_00_023_ALCOA_INC_Waiver_of_Domestic_and_Foreign_Patent_Ri.pdf U.S. Energy Requirements for Aluminum Production

  17. METHOD OF RECOVERING URANIUM COMPOUNDS

    DOE Patents [OSTI]

    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.

  18. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    2015 Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 thousand pounds U 3 O 8 equivalent Year Maximum ...

  19. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 2014 2015 2014 2015 2014 2015 Weighted-average price ...

  20. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    Figure 3. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by origin and delivery year, 2011-15 Figure 4. Weighted-average price of uranium ...

  1. Uranium-titanium-niobium alloy

    DOE Patents [OSTI]

    Ludtka, Gail M.; Ludtka, Gerard M.

    1990-01-01

    A uranium alloy having small additions of Ti and Nb shows improved strength and ductility in cross section of greater than one inch over prior uranium alloy having only Ti as an alloying element.

  2. METHOD OF SINTERING URANIUM DIOXIDE

    DOE Patents [OSTI]

    Henderson, C.M.; Stavrolakis, J.A.

    1963-04-30

    This patent relates to a method of sintering uranium dioxide. Uranium dioxide bodies are heated to above 1200 nif- C in hydrogen, sintered in steam, and then cooled in hydrogen. (AEC)

  3. Excess Uranium Inventory Management Plan

    Office of Energy Efficiency and Renewable Energy (EERE)

    The 2013 Excess Uranium Inventory Management Plan describes a framework for the effective management of the Energy Department’s surplus uranium inventory in support of meeting its critical...

  4. PROCESS FOR PRODUCING URANIUM HALIDES

    DOE Patents [OSTI]

    Murphree, E.V.

    1957-10-29

    A process amd associated apparatus for producing UF/sub 4/ from U/sub 3/ O/sub 8/ by a fluidized'' technique are reported. The U/sub 3/O/sub 8/ is first reduced to UO/sub 2/ by reaction with hydrogen, and the lower oxide of uranium is then reacted with gaseous HF to produce UF/sub 4/. In each case the reactant gas is used, alone or in combination with inert gases, to fluidize'' the finely divided reactant solid. The complete setup of the plant equipment including bins, reactor and the associated piping and valving, is described. An auxiliary fluorination reactor allows for the direct production of UF/sub 6/ from UF/sub 4/ and fluorine gas, or if desired, UF/sub 4/ may be collected as the product.

  5. uranium | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    uranium Klotz visits Y-12 to see progress on new projects and ongoing work on NNSA's national security missions Last week, NNSA Administrator Lt. Gen. Frank Klotz (Ret.) visited the Y-12 National Security Complex to check on the status of ongoing projects like the Uranium Processing Facility as well as the site's continuing uranium operations. He also met with the Region 2 volunteers of the Radiogical... NNSA Announces Arrival of Plutonium and Uranium from Japan's Fast Critical Assembly at

  6. Selective Extraction of Uranium from Liquid or Supercritical Carbon Dioxide

    SciTech Connect (OSTI)

    Farawila, Anne F.; O'Hara, Matthew J.; Wai, Chien M.; Taylor, Harry Z.; Liao, Yu-Jung

    2012-07-31

    Current liquid-liquid extraction processes used in recycling irradiated nuclear fuel rely on (1) strong nitric acid to dissolve uranium oxide fuel, and (2) the use of aliphatic hydrocarbons as a diluent in formulating the solvent used to extract uranium. The nitric acid dissolution process is not selective. It dissolves virtually the entire fuel meat which complicates the uranium extraction process. In addition, a solvent washing process is used to remove TBP degradation products, which adds complexity to the recycling plant and increases the overall plant footprint and cost. A liquid or supercritical carbon dioxide (l/sc -CO2) system was designed to mitigate these problems. Indeed, TBP nitric acid complexes are highly soluble in l/sc -CO2 and are capable of extracting uranium directly from UO2, UO3 and U3O8 powders. This eliminates the need for total acid dissolution of the irradiated fuel. Furthermore, since CO2 is easily recycled by evaporation at room temperature and pressure, it eliminates the complex solvent washing process. In this report, we demonstrate: (1) A reprocessing scheme starting with the selective extraction of uranium from solid uranium oxides into a TBP-HNO3 loaded Sc-CO2 phase, (2) Back extraction of uranium into an aqueous phase, and (3) Conversion of recovered purified uranium into uranium oxide. The purified uranium product from step 3 can be disposed of as low level waste, or mixed with enriched uranium for use in a reactor for another fuel cycle. After an introduction on the concept and properties of supercritical fluids, we first report the characterization of the different oxides used for this project. Our extraction system and our online monitoring capability using UV-Vis absorbance spectroscopy directly in sc-CO2 is then presented. Next, the uranium extraction efficiencies and kinetics is demonstrated for different oxides and under different physical and chemical conditions: l/sc -CO2 pressure and temperature, TBP/HNO3 complex used

  7. Fact #768: February 25, 2013 New Light Vehicle Sales and Gross Domestic

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Product | Department of Energy 8: February 25, 2013 New Light Vehicle Sales and Gross Domestic Product Fact #768: February 25, 2013 New Light Vehicle Sales and Gross Domestic Product Over the last four decades, new light vehicle sales have gone from a low of 9.9 million vehicles in 1970 to a high of 17.1 million vehicles sold in 2001, but along the way, there have been significant ups and downs. Those ups and downs are also reflected in the change in Gross Domestic Product (GDP) over time

  8. EXTRACTION OF URANIUM

    DOE Patents [OSTI]

    Kesler, R.D.; Rabb, D.D.

    1959-07-28

    An improved process is presented for recovering uranium from a carnotite ore. In the improved process U/sub 2/O/sub 5/ is added to the comminuted ore along with the usual amount of NaCl prior to roasting. The amount of U/sub 2/O/ sub 5/ is dependent on the amount of free calcium oxide and the uranium in the ore. Specifically, the desirable amount of U/sub 2/O/sub 5/ is 3.2% for each 1% of CaO, and 5 to 6% for each 1% of uranium. The mixture is roasted at about 1560 deg C for about 30 min and then leached with a 3 to 9% aqueous solution of sodium carbonate.

  9. Process for recovering uranium

    DOE Patents [OSTI]

    MacWood, G. E.; Wilder, C. D.; Altman, D.

    1959-03-24

    A process useful in recovering uranium from deposits on stainless steel liner surfaces of calutrons is presented. The deposit is removed from the stainless steel surface by washing with aqueous nitric acid. The solution obtained containing uranium, chromium, nickel, copper, and iron is treated with an excess of ammonium hydroxide to precipitnte the uranium, iron, and chromium and convert the nickel and copper to soluble ammonio complexions. The precipitated material is removed, dried and treated with carbon tetrachloride at an elevated temperature of about 500 to 600 deg C to form a vapor mixture of UCl/ sub 4/, UCl/sub 5/, FeCl/sub 3/, and CrCl/sub 4/. The UCl/sub 4/ is separated from this vapor mixture by selective fractional condensation at a temperature of about 500 to 400 deg C.

  10. Uranium industry annual, 1987

    SciTech Connect (OSTI)

    Not Available

    1988-09-29

    This report provides current statistical data on the US uranium industry for the Congress, federal and state agencies, the uranium and utility industries, and the public. It utilizes data from the mandatory ''Uranium Industry Annual Survey,'' Form EIA-858; historical data collected by the Energy Information Administration (EIA) and by the Grand Junction (Colorado) Project Office of the Idaho Operations Office of the US Department of Energy (DOE); and other data from federal agencies that preceded the DOE. The data provide a comprehensive statistical characterization of the industry's annual activities and include some information about industry plans and commitments over the next several years. Where these data are presented in aggregate form, care has been taken to protect the confidentiality of company-specific data while still conveying an accurate and complete statistical representation of the industry data.

  11. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    . Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by supplier and delivery year, 2011-15 thousand pounds U3O8 equivalent, dollars per pound U3O8 equivalent Deliveries 2011 2012 2013 2014 2015 Purchased from U.S. producers Purchases of U.S.-origin and foreign-origin uranium 550 W W W 1,455 Weighted-average price 58.12 W W W 52.35 Purchased from U.S. brokers and traders Purchases of U.S.-origin and foreign-origin uranium 14,778 11,545 12,835 17,111 13,852

  12. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    . Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by origin and delivery year, 2011-15 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries 2011 2012 2013 2014 2015 U.S.-Origin Uranium Purchases 5,205 9,807 9,484 3,316 3,419 Weighted-Average Price 52.12 59.44 56.37 48.11 43.86 Foreign-Origin Uranium Purchases 49,626 47,713 47,919 50,033 53,106 Weighted-Average Price 55.98 54.07 51.13 46.03 44.14 Total Purchases 54,831 57,520 57,403

  13. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    0. U.S. broker and trader purchases of uranium by origin, supplier, and delivery year, 2011-15 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries 2011 2012 2013 2014 2015 Received U.S.-origin uranium Purchases 1,668 1,194 W 410 2,702 Weighted-average price 54.85 51.78 W 33.55 35.04 Received foreign-origin uranium Purchases 24,695 24,606 W 28,743 33,014 Weighted-average price 49.69 47.75 W 38.42 39.58 Total received by U.S. brokers and traders Purchases 26,363 25,800

  14. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    1. Foreign sales of uranium from U.S. suppliers and owners and operators of U.S. civilian nuclear power reactors by origin and delivery year, 2011-15 thousands pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries to foreign suppliers and utilities 2011 2012 2013 2014 2015 U.S.-origin uranium Foreign sales 4,387 4,798 4,148 4,210 4,258 Weighted-average price 53.08 47.53 43.10 32.91 37.85 Foreign-origin uranium Foreign sales 12,297 13,185 14,717 15,794 21,465 Weighted-Average Price

  15. PROCESS FOR RECOVERING URANIUM

    DOE Patents [OSTI]

    MacWood, G.E.; Wilder, C.D.; Altman, D.

    1959-03-24

    A process is described for recovering uranium from deposits on stainless steel liner surfaces of calutrons. The deposit is removed from the stainless steel surface by washing with aqueous nitric acid. The solution obtained containing uranium, chromium, nickels copper, and iron is treated with excess of ammonium hydroxide to precipitatc the uranium, irons and chromium and convert thc nickel and copper to soluble ammonia complexions. The precipitated material is removed, dried, and treated with carbon tetrachloride at an elevated temperature of about 500 to 600 deg C to form a vapor mixture of UCl/sub 4/, UCl/sub 5/, FeCl/ sub 3/, and CrCl/sub 4/. The UCl/sub 4/ is separated from this vapor mixture by selective fractional condensation at a temprrature of about 300 to400 deg C.

  16. Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear...

    National Nuclear Security Administration (NNSA)

    the use of highly-enriched uranium (HEU) fuel to low-enriched uranium (LEU) as a key ... the use of highly-enriched uranium (HEU) fuel to low-enriched uranium (LEU) as a key ...

  17. PROCESS OF PREPARING URANIUM CARBIDE

    DOE Patents [OSTI]

    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)

  18. Uranium Transport Modeling

    SciTech Connect (OSTI)

    Bostick, William D.

    2008-01-15

    Uranium contamination is prevalent at many of the U.S. DOE facilities and at several civilian sites that have supported the nuclear fuel cycle. The potential off-site mobility of uranium depends on the partitioning of uranium between aqueous and solid (soil and sediment) phases. Hexavalent U (as uranyl, UO{sub 2}{sup 2+}) is relatively mobile, forming strong complexes with ubiquitous carbonate ion which renders it appreciably soluble even under mild reducing conditions. In the presence of carbonate, partition of uranyl to ferri-hydrate and select other mineral phases is usually maximum in the near-neutral pH range {approx} 5-8. The surface complexation reaction of uranyl with iron-containing minerals has been used as one means to model subsurface migration, used in conjunction with information on the site water chemistry and hydrology. Partitioning of uranium is often studied by short-term batch 'equilibrium' or long-term soil column testing ; MCLinc has performed both of these methodologies, with selection of method depending upon the requirements of the client or regulatory authority. Speciation of uranium in soil may be determined directly by instrumental techniques (e.g., x-ray photoelectron spectroscopy, XPS; x-ray diffraction, XRD; etc.) or by inference drawn from operational estimates. Often, the technique of choice for evaluating low-level radionuclide partitioning in soils and sediments is the sequential extraction approach. This methodology applies operationally-defined chemical treatments to selectively dissolve specific classes of macro-scale soil or sediment components. These methods recognize that total soil metal inventory is of limited use in understanding bioavailability or metal mobility, and that it is useful to estimate the amount of metal present in different solid-phase forms. Despite some drawbacks, the sequential extraction method can provide a valuable tool to distinguish among trace element fractions of different solubility related to

  19. METHOD OF ELECTROPOLISHING URANIUM

    DOE Patents [OSTI]

    Walker, D.E.; Noland, R.A.

    1959-07-14

    A method of electropolishing the surface of uranium articles is presented. The process of this invention is carried out by immersing the uranium anticle into an electrolyte which contains from 35 to 65% by volume sulfuric acid, 1 to 20% by volume glycerine and 25 to 50% by volume of water. The article is made the anode in the cell and polished by electrolyzing at a voltage of from 10 to 15 volts. Discontinuing the electrolysis by intermittently withdrawing the anode from the electrolyte and removing any polarized film formed therein results in an especially bright surface.

  20. TREATMENT OF URANIUM SURFACES

    DOE Patents [OSTI]

    Slunder, C.J.

    1959-02-01

    An improved process is presented for prcparation of uranium surfaces prior to electroplating. The surfacc of the uranium to be electroplated is anodized in a bath comprising a solution of approximately 20 to 602 by weight of phosphoric acid which contains about 20 cc per liter of concentrated hydrochloric acid. Anodization is carried out for approximately 20 minutes at a current density of about 0.5 amperes per square inch at a temperature of about 35 to 45 C. The oxidic film produced by anodization is removed by dipping in strong nitric acid, followed by rinsing with water just prior to electroplating.

  1. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    b. Weighted-average price of uranium purchased by owners and operators of U.S. civilian nuclear power reactors, 1994-2015 dollars per pound U3O8 equivalent Delivery year Total purchased (weighted-average price) Purchased from U.S. producers Purchased from U.S. brokers and traders Purchased from other owners and operators of U.S. civilian nuclear power reactors, other U.S. suppliers, (and U.S. government for 2007)1 Purchased from foreign suppliers U.S.-origin uranium (weighted-average price)

  2. PROCESS OF RECOVERING URANIUM

    DOE Patents [OSTI]

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

    1958-06-17

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

  3. Corrosion-resistant uranium

    DOE Patents [OSTI]

    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.

  4. Corrosion-resistant uranium

    DOE Patents [OSTI]

    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.

  5. Uranium Lease and Take-Back | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Uranium Lease and Take-Back

  6. URANIUM PRODUCERS OF AMERICA 141 EAST PALACE AVENUE, POST OFFICE Box 669, SANTA FE, NEW MEXICO 87504-0669

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AMERICA 141 EAST PALACE AVENUE, POST OFFICE Box 669, SANTA FE, NEW MEXICO 87504-0669 TELEPHONE(505) 982-4611; FAX (505) 988-2987; WWW.URANJUMPRODUCERSAMERJCA.COM David Henderson U.S. Department of Energy Office of Nuclear Energy Mail Stop NE-52 19901 Germantown Rd. Germantown, MD 20874-1290 April 6, 2015 Re: UP A Response to DOE Notice of Issues for Public Comment on "Excess Uranium Management: Effects of DOE Transfers of Excess Uranium on Domestic Uranium Mining, Conversion, and Enrichment

  7. High loading uranium fuel plate

    DOE Patents [OSTI]

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

    1990-01-01

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

  8. RECOVERY OF URANIUM FROM PITCHBLENDE

    DOE Patents [OSTI]

    Ruehle, A.E.

    1958-06-24

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

  9. STRIPPING OF URANIUM FROM ORGANIC EXTRACTANTS

    DOE Patents [OSTI]

    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)

  10. Uranium Reduction by Clostridia

    SciTech Connect (OSTI)

    Francis, A.J.; Dodge, Cleveland J.; Gillow, Jeffrey B.

    2006-04-05

    The FRC groundwater and sediment contain significant concentrations of U and Tc and are dominated by low pH, and high nitrate and Al concentrations where dissimilatory metal reducing bacterial activity may be limited. The presence of Clostridia in Area 3 at the FRC site has been confirmed and their ability to reduce uranium under site conditions will be determined. Although the phenomenon of uranium reduction by Clostridia has been firmly established, the molecular mechanisms underlying such a reaction are not very clear. The authors are exploring the hypothesis that U(VI) reduction occurs through hydrogenases and other enzymes (Matin and Francis). Fundamental knowledge of metal reduction using Clostridia will allow us to exploit naturally occurring processes to attenuate radionuclide and metal contaminants in situ in the subsurface. The outline for this report are as follows: (1) Growth of Clostridium sp. under normal culture conditions; (2) Fate of metals and radionuclides in the presence of Clostridia; (3) Bioreduction of uranium associated with nitrate, citrate, and lepidocrocite; and (4) Utilization of Clostridium sp. for immobilization of uranium at the FRC Area 3 site.

  11. DOE Selects Projects Totaling $12.4 Million Aimed at Increasing Domestic

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Energy Production While Enhancing Environmental Protection | Department of Energy Totaling $12.4 Million Aimed at Increasing Domestic Energy Production While Enhancing Environmental Protection DOE Selects Projects Totaling $12.4 Million Aimed at Increasing Domestic Energy Production While Enhancing Environmental Protection August 1, 2011 - 1:00pm Addthis Washington, DC - A total of 11 research projects that will help find ways to extract more energy from unconventional oil and gas resources

  12. Uranium Isotopic Assay Instrument

    SciTech Connect (OSTI)

    Anheier, Norman C.; Wojcik, Michael D.; Bushaw, Bruce A.

    2006-12-01

    The isotopic assay instrument under development at Pacific Northwest National Laboratory (PNNL) is capable of rapid prescreening to detect small and rare particles containing high concentrations of uranium in a heterogeneous sample. The isotopic measurement concept is based on laser vaporization of solid samples followed with sensitive isotope specific detection using either uranium atomic fluorescence emission or uranium atomic absorbance. Both isotopes are measured concurrently, following a single ablation laser pulse, using two external-cavity violet diode lasers. The simultaneous measurement of both isotopes enables the correlation of the fluorescence and absorbance signals on a shot-to-shot basis. This measurement approach demonstrated negligible channel crosstalk between isotopes. Rapid sample scanning provides high spatial resolution isotopic fluorescence and absorbance sample imagery of heterogeneous samples. Laser ablation combined with measurements of laser-induced fluorescence (LALIF) and through-plume laser absorbance (LAPLA) was applied to measure gadolinium isotope ratios in solid samples. Gadolinium has excitation wavelengths very close to the transitions of interest in uranium. Gadolinium has seven stable isotopes, and the natural 152Gd:160Gd ratio of 0.009 is in the range of what will be encountered for 235U:238U isotopic ratios. LAPLA measurements were demonstrated clearly using 152Gd (0.2% isotopic abundance) with a good signal-to-noise ratio. The ability to measure gadolinium abundances at this level indicates that measurements of 235U/238U isotopic ratios for natural (0.72%), depleted (0.25%), and low enriched uranium samples will be feasible.

  13. Occupational exposures to uranium: processes, hazards, and regulations

    SciTech Connect (OSTI)

    Stoetzel, G.A.; Fisher, D.R.; McCormack, W.D.; Hoenes, G.R.; Marks, S.; Moore, R.H.; Quilici, D.G.; Breitenstein, B.D.

    1981-04-01

    The United States Uranium Registry (USUR) was formed in 1978 to investigate potential hazards from occupational exposure to uranium and to assess the need for special health-related studies of uranium workers. This report provides a summary of Registry work done to date. The history of the uranium industry is outlined first, and the current commercial uranium industry (mining, milling, conversion, enrichment, and fuel fabrication) is described. This description includes information on basic processes and areas of greatest potential radiological exposure. In addition, inactive commercial facilities and other uranium operations are discussed. Regulation of the commercial production industry for uranium fuel is reported, including the historic development of regulations and the current regulatory agencies and procedures for each phase of the industry. A review of radiological health practices in the industry - facility monitoring, exposure control, exposure evaluation, and record-keeping - is presented. A discussion of the nonradiological hazards of the industry is provided, and the final section describes the tissue program developed as part of the Registry.

  14. Secure Fuels from Domestic Resources - Oil Shale and Tar Sands...

    Office of Environmental Management (EM)

    Secure Fuels from Domestic Resources - Oil Shale and Tar Sands Secure Fuels from Domestic Resources - Oil Shale and Tar Sands Profiles of Companies Engaged in Domestic Oil Shale ...

  15. Reaction of uranium oxides with chlorine and carbon or carbon monoxide to prepare uranium chlorides

    SciTech Connect (OSTI)

    Haas, P.A.; Lee, D.D.; Mailen, J.C.

    1991-11-01

    The preferred preparation concept of uranium metal for feed to an AVLIS uranium enrichment process requires preparation of uranium tetrachloride (UCI{sub 4}) by reacting uranium oxides (UO{sub 2}/UO{sub 3}) and chlorine (Cl{sub 2}) in a molten chloride salt medium. UO{sub 2} is a very stable metal oxide; thus, the chemical conversion requires both a chlorinating agent and a reducing agent that gives an oxide product which is much more stable than the corresponding chloride. Experimental studies in a quartz reactor of 4-cm ID have demonstrated the practically of some chemical flow sheets. Experimentation has illustrated a sequence of results concerning the chemical flow sheets. Tests with a graphite block at 850{degrees}C demonstrated rapid reactions of Cl{sub 2} and evolution of carbon dioxide (CO{sub 2}) as a product. Use of carbon monoxide (CO) as the reducing agent also gave rapid reactions of Cl{sub 2} and formation of CO{sub 2} at lower temperatures, but the reduction reactions were slower than the chlorinations. Carbon powder in the molten salt melt gave higher rates of reduction and better steady state utilization of Cl{sub 2}. Addition of UO{sub 2} feed while chlorination was in progress greatly improved the operation by avoiding the plugging effects from high UO{sub 2} concentrations and the poor Cl{sub 2} utilizations from low UO{sub 2} concentrations. An UO{sub 3} feed gave undesirable effects while a feed of UO{sub 2}-C spheres was excellent. The UO{sub 2}-C spheres also gave good rates of reaction as a fixed bed without any molten chloride salt. Results with a larger reactor and a bottom condenser for volatilized uranium show collection of condensed uranium chlorides as a loose powder and chlorine utilizations of 95--98% at high feed rates. 14 refs., 7 figs., 14 tabs.

  16. 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

    DOE Patents [OSTI]

    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.

  17. 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

    DOE Patents [OSTI]

    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.

  18. Assessment of Unglazed Solar Domestic Water Heaters

    SciTech Connect (OSTI)

    Burch, J.; Salasovich, J.; Hillman, T.

    2005-12-01

    Conference paper investigating cost-performance tradeoffs in replacing glazed collectors with unglazed collectors in solar domestic water heating systems.

  19. U.S. Energy Information Administration (EIA) - Ap

    U.S. Energy Information Administration (EIA) Indexed Site

    Domestic Uranium Production Report - Quarterly Information on uranium production and the number of producing facilities Consumption & Efficiency view all Residential Energy ...

  20. U.S. Energy Information Administration (EIA) - Ap

    Gasoline and Diesel Fuel Update (EIA)

    Status of U.S. Nuclear Outages Daily statistics on nuclear capacity by plant. Domestic Uranium Production Report - Quarterly Information on uranium production and the number of ...

  1. Uranium-series constraints on radionuclide transport and groundwater flow at the Nopal I uranium deposit, Sierra Pena Blanca, Mexico

    SciTech Connect (OSTI)

    Goldstein, S.J.; Abdel-Fattah, A.I.; Murrell, M.T.; Dobson, P.F.; Norman, D.E.; Amato, R.S.; Nunn, A. J.

    2009-10-01

    Uranium-series data for groundwater samples from the Nopal I uranium ore deposit were obtained to place constraints on radionuclide transport and hydrologic processes for a nuclear waste repository located in fractured, unsaturated volcanic tuff. Decreasing uranium concentrations for wells drilled in 2003 are consistent with a simple physical mixing model that indicates that groundwater velocities are low ({approx}10 m/y). Uranium isotopic constraints, well productivities, and radon systematics also suggest limited groundwater mixing and slow flow in the saturated zone. Uranium isotopic systematics for seepage water collected in the mine adit show a spatial dependence which is consistent with longer water-rock interaction times and higher uranium dissolution inputs at the front adit where the deposit is located. Uranium-series disequilibria measurements for mostly unsaturated zone samples indicate that {sup 230}Th/{sup 238}U activity ratios range from 0.005-0.48 and {sup 226}Ra/{sup 238}U activity ratios range from 0.006-113. {sup 239}Pu/{sup 238}U mass ratios for the saturated zone are <2 x 10{sup -14}, and Pu mobility in the saturated zone is >1000 times lower than the U mobility. Saturated zone mobility decreases in the order {sup 238}U{approx}{sup 226}Ra > {sup 230}Th{approx}{sup 239}Pu. Radium and thorium appear to have higher mobility in the unsaturated zone based on U-series data from fractures and seepage water near the deposit.

  2. Method of preparation of uranium nitride

    DOE Patents [OSTI]

    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.

  3. Matrix Infrared Spectroscopic and Computational Investigations of Novel Small Uranium Containing Molecules - Final Technical Report

    SciTech Connect (OSTI)

    Andrews, Lester

    2014-10-17

    Direct reactions of f-element uranium, thorium and lanthanide metal atoms were investigated with small molecules. These metal atoms were generated by laser ablation and mixed with the reagent molecules then condensed with noble gases at 4K. The products were analyzed by absorption of infrared light to measure vibrational frequencies which were confirmed by quantum chemical calculations. We have learned more about the reactivity of uranium atoms with common molecules, which will aid in the develolpment of further applications of uranium.

  4. Survey of lands held for uranium exploration, development, and production in fourteen western states for the six-month period ending December 31, 1982. [AZ, CA, CO, ID, MT, NV, NM, ND, OR, SD, TX, UT, WA, WY

    SciTech Connect (OSTI)

    Not Available

    1983-04-01

    Tables are presented showing the land held (in thousands of acres) for uranium exploration and development according to: (1) distribution by ownership for 14 western states; (2) distribution by state; distribution by land category. A graph is presented showing land held by uranium industry from January 1966 to January 1983. Land controlled by counties for each of the 14 states are also tabulated according to land category. (ATT)

  5. Survey of lands held for uranium exploration, development, and production in fourteen western states for the six-month period ending June 30, 1982. [AZ, CA, CO, ID, MT, NV, NM, ND, OR, SD, TX, UT, WA, WY

    SciTech Connect (OSTI)

    Not Available

    1982-10-01

    Tables are presented showing the land held (in thousands of acres) for uranium exploration according to: (1) distribution by ownership for 14 western states (state, claim, federally acquired, Indian and fee); (2) distribution by state (1976 to 1982); (3) distribution by land category (1976 to 1982). A graph is presented showing land held by uranium industry from January 1966 to January 1982. Land controlled by counties for each of the 14 states are also tabulated according to land category. (ATT)

  6. file://\\fs-f1\shared\uranium\uranium.html

    U.S. Energy Information Administration (EIA) Indexed Site

    Glossary Home > Nuclear > U.S. Uranium Reserves Estimates U.S. Uranium Reserves Estimates Data for: 2008 Report Released: July 2010 Next Release Date: 2012 Summary The U.S. Energy Information Administration (EIA) has updated its estimates of uranium reserves for year-end 2008. This represents the first revision of the estimates since 2004. The update is based on analysis of company annual reports, any additional information reported by companies at conferences and in news releases,

  7. Method of preparing uranium nitride or uranium carbonitride bodies

    DOE Patents [OSTI]

    Wilhelm, Harley A.; McClusky, James K.

    1976-04-27

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

  8. Aluminum and polymeric coatings for protection of uranium

    SciTech Connect (OSTI)

    Colmenares, C.; McCreary, T.; Monaco, S.; Walkup, C.; Gleeson, G.; Kervin, J.; Smith, R.L.; McCaffrey, C.

    1983-12-21

    Ion-plated aluminum films on uranium will not provide adequate protection for 25 years. Magnetron-plated aluminum films on uranium are much better than ion-plated ones. Kel-F 800 films on uranium can provide adequate protection for 25 years. Their use in production must be delayed until the following factors are sorted out: water permeability in Kel-F 800 must be determined between 30 and 60/sup 0/C; the effect of UF/sub 3/, at the Kel-F/metal interface, on the permeability of water must be assessed; and the effect of crystallinity on water permeability must be evaluated. Applying Kel-F films on aluminum ion-plated uranium provides a good interim solution for long term storage.

  9. RECOVERY OF URANIUM FROM ZIRCONIUM-URANIUM NUCLEAR FUELS

    DOE Patents [OSTI]

    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)

  10. PREPARATION OF DENSE URANIUM DIOXIDE PARTICLES FROM URANIUM HEXAFLUORI...

    Office of Scientific and Technical Information (OSTI)

    Visit OSTI to utilize additional information resources in energy science and technology. A ... A fluid-bed method was developed for the direct preparation from uranium hexafluoride of ...

  11. Method for fabricating uranium foils and uranium alloy foils

    DOE Patents [OSTI]

    Hofman, Gerard L.; Meyer, Mitchell K.; Knighton, Gaven C.; Clark, Curtis R.

    2006-09-05

    A method of producing thin foils of uranium or an alloy. The uranium or alloy is cast as a plate or sheet having a thickness less than about 5 mm and thereafter cold rolled in one or more passes at substantially ambient temperatures until the uranium or alloy thereof is in the shape of a foil having a thickness less than about 1.0 mm. The uranium alloy includes one or more of Zr, Nb, Mo, Cr, Fe, Si, Ni, Cu or Al.

  12. PROCESS FOR PRODUCING URANIUM TETRAFLUORIDE

    DOE Patents [OSTI]

    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.

  13. WELDED JACKETED URANIUM BODY

    DOE Patents [OSTI]

    Gurinsky, D.H.

    1958-08-26

    A fuel element is presented for a neutronic reactor and is comprised of a uranium body, a non-fissionable jacket surrounding sald body, thu jacket including a portion sealed by a weld, and an inclusion in said sealed jacket at said weld of a fiux having a low neutron capture cross-section. The flux is provided by combining chlorine gas and hydrogen in the intense heat of-the arc, in a "Heliarc" welding muthod, to form dry hydrochloric acid gas.

  14. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    b. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by purchaser, 2013-15 deliveries thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries in 2013 Deliveries in 2014 Deliveries in 2015 Distribution of purchasers Number of purchasers Quantity with reported price Weighted-average price Number of purchasers Quantity with reported price Weighted-average price Number of purchasers Quantity with reported price

  15. Selective leaching of uranium from uranium-contaminated soils: Progress report 1

    SciTech Connect (OSTI)

    Francis, C.W.; Mattus, A.J.; Farr, L.L.; Elless, M.P.; Lee, S.Y.

    1993-02-01

    Three soils and a sediment contaminated with uranium were used to determine the effectiveness of sodium carbonate and citric acid leaching to decontaminated or remove uranium to acceptable regulatory levels. Two of the soils were surface soils from the DOE facility formerly called the Feed Materials Production Center (FMPC) at Fernald, Ohio. This facility is presently called the Femald Environmental Management Project (FEMP). Carbonate extractions generally removed from 70 to 90% of the uranium from the Fernald storage pad soil. Uranium was slightly more difficult to extract from the Fernald incinerator and the Y-12 landfarm soils. Very small amounts of uranium could be extracted from the storm sewer sediment. Extraction with carbonate at high solution-to-soil ratios were as effective as extractions at low solution-to-soil ratios, indicating attrition by the paddle mixer was not significantly different than that provided in a rotary extractor. Also, pretreatments such as milling or pulverizing the soil sample did not appear to increase extraction efficiency when carbonate extractions were carried out at elevated temperatures (60{degree}C) or long extraction times (23 h). Adding KMnO{sub 4} in the carbonate extraction appeared to increase extraction efficiency from the Fernald incinerator soil but not the Fernald storage pad soil. The most effective leaching rates (> 90 % from both Fernald soils) were obtained using a citrate/dithionite extraction procedure designed to remove amorphous (noncrystalline) iron/aluminum sesquioxides from surfaces of clay minerals. Citric acid also proved to be a very good extractant for uranium.

  16. Selective leaching of uranium from uranium-contaminated soils: Progress report 1

    SciTech Connect (OSTI)

    Francis, C.W.; Mattus, A.J.; Farr, L.L.; Elless, M.P.; Lee, S.Y.

    1993-02-01

    Three soils and a sediment contaminated with uranium were used to determine the effectiveness of sodium carbonate and citric acid leaching to decontaminated or remove uranium to acceptable regulatory levels. Two of the soils were surface soils from the DOE facility formerly called the Feed Materials Production Center (FMPC) at Fernald, Ohio. This facility is presently called the Femald Environmental Management Project (FEMP). Carbonate extractions generally removed from 70 to 90% of the uranium from the Fernald storage pad soil. Uranium was slightly more difficult to extract from the Fernald incinerator and the Y-12 landfarm soils. Very small amounts of uranium could be extracted from the storm sewer sediment. Extraction with carbonate at high solution-to-soil ratios were as effective as extractions at low solution-to-soil ratios, indicating attrition by the paddle mixer was not significantly different than that provided in a rotary extractor. Also, pretreatments such as milling or pulverizing the soil sample did not appear to increase extraction efficiency when carbonate extractions were carried out at elevated temperatures (60[degree]C) or long extraction times (23 h). Adding KMnO[sub 4] in the carbonate extraction appeared to increase extraction efficiency from the Fernald incinerator soil but not the Fernald storage pad soil. The most effective leaching rates (> 90 % from both Fernald soils) were obtained using a citrate/dithionite extraction procedure designed to remove amorphous (noncrystalline) iron/aluminum sesquioxides from surfaces of clay minerals. Citric acid also proved to be a very good extractant for uranium.

  17. METHOD OF DISSOLVING URANIUM METAL

    DOE Patents [OSTI]

    Slotin, L.A.

    1958-02-18

    This patent relates to an economicai means of dissolving metallic uranium. It has been found that the addition of a small amount of perchloric acid to the concentrated nitric acid in which the uranium is being dissolved greatly shortens the time necessary for dissolution of the metal. Thus the use of about 1 or 2 percent of perchioric acid based on the weight of the nitric acid used, reduces the time of dissolution of uranium by a factor of about 100.

  18. PROCESS FOR PREPARING URANIUM METAL

    DOE Patents [OSTI]

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

    1959-01-13

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

  19. SEPARATION OF THORIUM FROM URANIUM

    DOE Patents [OSTI]

    Bane, R.W.

    1959-09-01

    A description is given for the separation of thorium from uranium by forming an aqueous acidic solution containing ionic species of thorium, uranyl uranium, and hydroxylamine, flowing the solution through a column containing the phenol-formaldehyde type cation exchange resin to selectively adsorb substantially all the thorium values and a portion of the uranium values, flowing a dilute solution of hydrochloric acid through the column to desorb the uranium values, and then flowing a dilute aqueous acidic solution containing an ion, such as bisulfate, which has a complexing effect upon thortum through the column to desorb substantially all of the thorium.

  20. Highly Enriched Uranium Materials Facility

    National Nuclear Security Administration (NNSA)

    Appropriations Subcommittee, is shown some of the technology in the Highly Enriched Uranium Materials Facility by Warehousing and Transportation Operations Manager Byron...

  1. 2014 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    By law, EIA's data, analyses, and forecasts are independent ... on information reported on Form EIA-858, "Uranium Marketing ... nuclear power reactors by contract type and material type, ...

  2. THERMAL DECOMPOSITION OF URANIUM COMPOUNDS

    DOE Patents [OSTI]

    Magel, T.T.; Brewer, L.

    1959-02-10

    A method is presented of preparing uranium metal of high purity consisting contacting impure U metal with halogen vapor at between 450 and 550 C to form uranium halide vapor, contacting the uranium halide vapor in the presence of H/sub 2/ with a refractory surface at about 1400 C to thermally decompose the uranium halides and deposit molten U on the refractory surface and collecting the molten U dripping from the surface. The entire operation is carried on at a sub-atmospheric pressure of below 1 mm mercury.

  3. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    3 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Quantity with reported price Weighted-average price Quantity with reported price ...

  4. 2015 Uranium Market Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    received in 2015","Weighted-average price","Number of purchase contracts for ... Administration, Form EIA-858 ""Uranium Marketing Annual Survey"" (2015)." "16 ...

  5. 2015 Uranium Marketing Annual Survey

    U.S. Energy Information Administration (EIA) Indexed Site

    5 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Quantity with reported price Weighted-average price Quantity with reported price ...

  6. METHOD FOR RECOVERING URANIUM FROM OILS

    DOE Patents [OSTI]

    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.

  7. Uranium Oxide Aerosol Transport in Porous Graphite

    SciTech Connect (OSTI)

    Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A.; Bagaasen, Larry M.; Brown, Charles C.; Iovin, Cristian; Delegard, Calvin H.; Zelenyuk, Alla; Buck, Edgar C.; Riley, Brian J.; Burns, Carolyn A.

    2012-01-23

    The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactors lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

  8. Depleted and Recyclable Uranium in the United States: Inventories and Options

    SciTech Connect (OSTI)

    Schneider, Erich; Scopatza, Anthony; Deinert, Mark

    2007-07-01

    International consumption of uranium currently outpaces production by nearly a factor of two. Secondary supplies from dismantled nuclear weapons, along with civilian and governmental stockpiles, are being used to make up the difference but supplies are limited. Large amounts of {sup 235}U are contained in spent nuclear fuel as well as in the tails left over from past uranium enrichment. The usability of these inhomogeneous uranium supplies depends on their isotopics. We present data on the {sup 235}U content of spent nuclear fuel and depleted uranium tails in the US and discuss the factors that affect its marketability and alternative uses. (authors)

  9. Uranium Enrichment Decontamination and Decommissioning Fund's...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2008 and 2007 Financial Statement Audit, OAS-FS-10-05 Uranium Enrichment Decontamination and...

  10. Calculating Atomic Number Densities for Uranium

    Energy Science and Technology Software Center (OSTI)

    1993-01-01

    Provides method to calculate atomic number densities of selected uranium compounds and hydrogenous moderators for use in nuclear criticality safety analyses at gaseous diffusion uranium enrichment facilities.

  11. Nuclear radiation cleanup and uranium prospecting (Patent) |...

    Office of Scientific and Technical Information (OSTI)

    Nuclear radiation cleanup and uranium prospecting Citation Details In-Document Search Title: Nuclear radiation cleanup and uranium prospecting Apparatus, systems, and methods for...

  12. Nuclear radiation cleanup and uranium prospecting (Patent) |...

    Office of Scientific and Technical Information (OSTI)

    Nuclear radiation cleanup and uranium prospecting Citation Details In-Document Search Title: Nuclear radiation cleanup and uranium prospecting You are accessing a document from...

  13. Uranium Resources Inc URI | Open Energy Information

    Open Energy Info (EERE)

    exploring, developing and mining uranium properties using the in situ recovery (ISR) or solution mining process. References: Uranium Resources, Inc. (URI)1 This article...

  14. Uranium Biomineralization By Natural Microbial Phosphatase Activities...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Uranium Biomineralization By Natural Microbial Phosphatase Activities in the Subsurface Citation Details In-Document Search Title: Uranium Biomineralization By ...

  15. PROCESS OF PRODUCING REFRACTORY URANIUM OXIDE ARTICLES

    DOE Patents [OSTI]

    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.

  16. Structural Sequestration of Uranium in Bacteriogenic Manganese...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Sequestration of Uranium in Bacteriogenic Manganese Oxides Samuel M. Webb (Stanford ... Uranium is a key contaminant of concern at US DOE sites and shuttered mining and ore ...

  17. Uranium Processing Facility team signs partnering agreement ...

    National Nuclear Security Administration (NNSA)

    Uranium Processing Facility team signs partnering agreement Thursday, July 24, 2014 - 9:40am Officials from NNSA's Uranium Processing Facility Project Office and Consolidated ...

  18. Conversion of depleted uranium hexafluoride to a solid uranium compound

    DOE Patents [OSTI]

    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.

  19. Novel Sensor for the In Situ Measurement of Uranium Fluxes

    SciTech Connect (OSTI)

    Hatfield, Kirk

    2015-02-10

    The goal of this project was to develop a sensor that incorporates the field-tested concepts of the passive flux meter to provide direct in situ measures of flux for uranium and groundwater in porous media. Measurable contaminant fluxes [J] are essentially the product of concentration [C] and groundwater flux or specific discharge [q ]. The sensor measures [J] and [q] by changes in contaminant and tracer amounts respectively on a sorbent. By using measurement rather than inference from static parameters, the sensor can directly advance conceptual and computational models for field scale simulations. The sensor was deployed in conjunction with DOE in obtaining field-scale quantification of subsurface processes affecting uranium transport (e.g., advection) and transformation (e.g., uranium attenuation) at the Rifle IFRC Site in Rifle, Colorado. Project results have expanded our current understanding of how field-scale spatial variations in fluxes of uranium, groundwater and salient electron donor/acceptors are coupled to spatial variations in measured microbial biomass/community composition, effective field-scale uranium mass balances, attenuation, and stability. The coupling between uranium, various nutrients and micro flora can be used to estimate field-scale rates of uranium attenuation and field-scale transitions in microbial communities. This research focuses on uranium (VI), but the sensor principles and design are applicable to field-scale fate and transport of other radionuclides. Laboratory studies focused on sorbent selection and calibration, along with sensor development and validation under controlled conditions. Field studies were conducted at the Rifle IFRC Site in Rifle, Colorado. These studies were closely coordinated with existing SBR (formerly ERSP) projects to complement data collection. Small field tests were conducted during the first two years that focused on evaluating field-scale deployment procedures and validating sensor performance under

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

    DOE Patents [OSTI]

    Visser, Ann E; Bridges, Nicholas J

    2014-06-10

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

  1. S. 210: This Act may be referred to as the Comprehensive Uranium Act of 1991, introduced in the Senate of the United States, One Hundred Second Congress, First Session, January 15, 1991

    SciTech Connect (OSTI)

    Not Available

    1991-01-01

    This bill would establish the United States Enrichment Corporation to operate the Federal uranium enrichment program on a profitable and efficient basis in order to maximize the long term economic value to the US, provide assistance to the domestic uranium industry, and provide a Federal contribution for the reclamation of mill tailings generated pursuant to Federal defense contracts at active uranium and thorium processing sites. The bill describes congressional findings; definitions, establishment of corporation, and purposes; corporate offices; powers and duties of the corporation; organization, finance, and management; licensing, taxation, and miscellaneous provisions; decontamination and decommissioning; and uranium security and tailing reclamation.

  2. Uranium exploration of the Colorado Plateau: interim staff report

    SciTech Connect (OSTI)

    Not Available

    1980-10-01

    This report is an issue of the original draft copy of the Interim Staff Report on Uranium Exploration on the Colorado Plateau, dated June 1951. The original draft copy was only recently located and is being published at this time because of the interest in the contained historical content. The table of contents of this report lists: history of uranium mining; geology; proposed program for the geologic investigations section; general activities of industry and government; and future exploration of sedimentary uranium deposits and anticipated results. Under the proposed program section are: future of the copper-uranium deposits as a source of uranium; uraniferous asphaltite deposits; and commission exploration and future possibilities. The section on general activities of industry and government includes: exploratory and development drilling; field investigations and mapping; early geologic investigations and investigations by the US geological survey; and geophysical exploration. Tables are also presented on: uranium production by districts; US Geological survey drilling statistics; Colorado Exploration Branch drilling statistics; summary of drilling projects; and comparative yearly core-drill statistics on the Colorado Plateau.

  3. Uranium Biomineralization by Natural Microbial Phosphatase Activities in the Subsurface

    SciTech Connect (OSTI)

    Sobecky, Patricia A.

    2015-04-06

    In this project, inter-disciplinary research activities were conducted in collaboration among investigators at The University of Alabama (UA), Georgia Institute of Technology (GT), Lawrence Berkeley National Laboratory (LBNL), Brookhaven National Laboratory (BNL), the DOE Joint Genome Institute (JGI), and the Stanford Synchrotron Radiation Light source (SSRL) to: (i) confirm that phosphatase activities of subsurface bacteria in Area 2 and 3 from the Oak Ridge Field Research Center result in solid U-phosphate precipitation in aerobic and anaerobic conditions; (ii) investigate the eventual competition between uranium biomineralization via U-phosphate precipitation and uranium bioreduction; (iii) determine subsurface microbial community structure changes of Area 2 soils following organophosphate amendments; (iv) obtain the complete genome sequences of the Rahnella sp. Y9-602 and the type-strain Rahnella aquatilis ATCC 33071 isolated from these soils; (v) determine if polyphosphate accumulation and phytate hydrolysis can be used to promote U(VI) biomineralization in subsurface sediments; (vi) characterize the effect of uranium on phytate hydrolysis by a new microorganism isolated from uranium-contaminated sediments; (vii) utilize positron-emission tomography to label and track metabolically-active bacteria in soil columns, and (viii) study the stability of the uranium phosphate mineral product. Microarray analyses and mineral precipitation characterizations were conducted in collaboration with DOE SBR-funded investigators at LBNL. Thus, microbial phosphorus metabolism has been shown to have a contributing role to uranium immobilization in the subsurface.

  4. SOLVENT EXTRACTION OF URANIUM VALUES

    DOE Patents [OSTI]

    Feder, H.M.; Ader, M.; Ross, L.E.

    1959-02-01

    A process is presented for extracting uranium salt from aqueous acidic solutions by organic solvent extraction. It consists in contacting the uranium bearing solution with a water immiscible dialkylacetamide having at least 8 carbon atoms in the molecule. Mentioned as a preferred extractant is dibutylacetamide. The organic solvent is usually used with a diluent such as kerosene or CCl/sub 4/.

  5. PLUTONIUM-URANIUM-TITANIUM ALLOYS

    DOE Patents [OSTI]

    Coffinberry, A.S.

    1959-07-28

    A plutonium-uranium alloy suitable for use as the fuel element in a fast breeder reactor is described. The alloy contains from 15 to 60 at.% titanium with the remainder uranium and plutonium in a specific ratio, thereby limiting the undesirable zeta phase and rendering the alloy relatively resistant to corrosion and giving it the essential characteristic of good mechanical workability.

  6. ELECTRODEPOSITION OF NICKEL ON URANIUM

    DOE Patents [OSTI]

    Gray, A.G.

    1958-08-26

    A method is described for preparing uranium objects prior to nickel electroplating. The process consiats in treating the surface of the uranium with molten ferric chloride hexahydrate, at a slightiy elevated temperature. This treatment etches the metal surface providing a structure suitable for the application of adherent electrodeposits and at the same time plates the surface with a thin protective film of iron.

  7. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    3. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by origin country and delivery year, 2011-15 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries in 2011 Deliveries in 2012 Deliveries in 2013 Deliveries in 2014 Deliveries in 2015 Origin country Purchases Weighted-average price Purchases Weighted-average price Purchases Weighted-average price Purchases Weighted-average price Purchases Weighted-average price Australia 6,001 57.47 6,724

  8. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    a. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by quantity, 2013-15 deliveries thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries in 2013 Deliveries in 2014 Deliveries in 2015 Quantity 1 distribution Quantity with reported price Weighted-average price Quantity with reported price Weighted-average price Quantity with reported price Weighted-average price First 7,175 34.34 6,665 30.26 6,807 29.68

  9. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    7. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by contract type and material type, 2015 deliveries thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Spot 1 Contracts Long-Term Contracts 2 Total Material Type Quantity with reported price Weighted-average price Quantity with reported price Weighted-average price Quantity with reported price Weighted-average price U3O8 6,175 36.40 24,107 45.76 30,282 43.85 Natural UF6 3,879 38.52 12,292 48.13

  10. DOE/NNSA Successfully Establishes Uranium Lease and Takeback Program to

    National Nuclear Security Administration (NNSA)

    Support Critical Medical Isotope Production | National Nuclear Security Administration | (NNSA) Successfully Establishes Uranium Lease and Takeback Program to Support Critical Medical Isotope Production Tuesday, February 16, 2016 - 12:00am In January 2016, the U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA) successfully established the Uranium Lease and Take-Back (ULTB) program, as directed in the American Medical Isotopes Production Act of 2012, to support

  11. METHOD OF ELECTROPLATING ON URANIUM

    DOE Patents [OSTI]

    Rebol, E.W.; Wehrmann, R.F.

    1959-04-28

    This patent relates to a preparation of metallic uranium surfaces for receiving coatings, particularly in order to secure adherent electroplated coatings upon uranium metal. In accordance with the invention the uranium surface is pretreated by degreasing in trichloroethylene, followed by immersion in 25 to 50% nitric acid for several minutes, and then rinsed with running water, prior to pickling in trichloroacetic acid. The last treatment is best accomplished by making the uranium the anode in an aqueous solution of 50 per cent by weight trichloroacetic acid until work-distorted crystals or oxide present on the metal surface have been removed and the basic crystalline structure of the base metal has been exposed. Following these initial steps the metallic uranium is rinsed in dilute nitric acid and then electroplated with nickel. Adnerent firmly-bonded coatings of nickel are obtained.

  12. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

    Friedman, Horace A.

    1985-01-01

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

  13. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

    Friedman, H.A.

    1984-06-13

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

  14. Domestic Nuclear Detection Office's Approach to Detect

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Domestic Nuclear Detection Office's Approach to Detect Concealed Threats Joel Rynes, PhD Department of Homeland Security (DHS) Domestic Nuclear Detection Office (DNDO) August 5, 2015 4:00 p.m. The Transformational and Applied Research (TAR) Directorate within the Domestic Nuclear Detection Office (DNDO) of the Department of Homeland Security (DHS) has the mission to develop break-through technologies that will have a dramatic impact on capabilities to detect nuclear and radiological threats

  15. Smart Domestic Appliances Provide Flexibility for Sustainable...

    Open Energy Info (EERE)

    URI: cleanenergysolutions.orgcontentsmart-domestic-appliances-provide-fle Language: English Policies: "Deployment Programs,Regulations" is not in the list of possible...

  16. Seeking New Approaches to Investigate Domestication Events |...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Seeking New Approaches to Investigate Domestication Events Monday, October 29, 2012 - 3:30am SSRL Bldg. 137, Rm. 322 Krish Seetah, Stanford University, Department of Anthropology...

  17. THE RECOVERY OF URANIUM FROM GAS MIXTURE

    DOE Patents [OSTI]

    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)

  18. Process for removing carbon from uranium

    DOE Patents [OSTI]

    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.

  19. PREPARATION OF URANIUM-ALUMINUM ALLOYS

    DOE Patents [OSTI]

    Moore, R.H.

    1962-09-01

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

  20. Uranium enrichment: investment options for the long term

    SciTech Connect (OSTI)

    Not Available

    1983-01-01

    The US government supplies a major portion of the enriched uranium used to fuel most of the nuclear power plants that furnish electricity in the free world. As manager of the US uranium enrichment concern, the Department of Energy (DOE) is investigating a number of technological choices to improve enrichment service and remain a significant world supplier. The Congress will ultimately select a strategy for federal investment in the uranium enrichment enterprise. A fundamental policy choice between possible future roles - that of the free world's main supplier of enrichment services, and that of a mainly domestic supplier - will underlie any investment decision the Congress makes. The technological choices are gaseous diffusion, gas centrifuge, and atomic vapor laser isotope separation (AVLIS). A base plan and four alternatives were examined by DOE and the Congressional Budget Office. In terms of total enterprise costs, Option IV, ultimately relying on advanced gas centrifuges for enrichment services, would offer the most economic approach, with costs over the full projection period totaling $123.5 billion. Option III, ultimately relying on AVLIS without gas centrifuge enrichment or gaseous diffusion, falls next in the sequence, with costs of $128.2 billion. Options I and II, involving combinations of the gas centrifuge and AVLIS technologies, follow closely with costs of $128.7 and $129.6 billion. The base plan has costs of $136.8 billion over the projection period. 1 figure, 22 tables.

  1. Prices dip, activity increases in unrestricted uranium market. [Uranium market overview

    SciTech Connect (OSTI)

    Not Available

    1993-05-01

    April's activity in the restricted uranium market fluctuated in the same range as that observed in March. At the same time, NUKEM detects a weakening of prices in the unrestricted market to $7.45-$7.65. Unrestricted buyers seem to have detected lower prices as well; much of the new demand noted this month emerged in the unrestricted segment of the market. With this issue, NUKEM inaugurates a new market statistic. To better follow developments in the conversion market, we will report a spot price range for conversion services. This price measure will be derived in a manner analogous to NUKEM's other spot market price ranges. We will continue to publish the current NUKEM price range for new contracts for a few months. If you wish to retain the old conversion contract price range in future editions, please contact our US office. Four deals for near term delivery occurred in the uranium market in April, resulting in spot market transaction volume of 2.5 million lbs U3O8 equivalent. In the first week, a US non-utility purchased a small quantity of enriched uranium product from an intermediary in a spot transaction representing about 75,000 lbs U3O8. The second week saw the stealthy purchase of Portland General Electric's inventory of natural and enriched uranium. The buyer of PGE's 1.1 million lbs U3O8 equivalent has achieved an unusual degree of anonymity. Also during the second week, a US utility bought a small quantity of enriched uranium containing less than 25,000 lbs natural U3O8 equivalent.

  2. ELUTION OF URANIUM FROM RESIN

    DOE Patents [OSTI]

    McLEan, D.C.

    1959-03-10

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

  3. Evaluation of a RF-Based Approach for Tracking UF6 Cylinders at a Uranium Enrichment Plant

    SciTech Connect (OSTI)

    Pickett, Chris A; Younkin, James R; Kovacic, Donald N; Laughter, Mark D; Hines, Jairus B; Boyer, Brian; Martinez, B.

    2008-01-01

    Approved industry-standard cylinders are used globally to handle and store uranium hexafluoride (UF{sub 6}) feed, product, tails, and samples at uranium enrichment plants. The International Atomic Energy Agency (IAEA) relies on time-consuming physical inspections to verify operator declarations and detect possible diversion of UF{sub 6}. Development of a reliable, automated, and tamper-resistant system for near real-time tracking and monitoring UF{sub 6} cylinders (as they move within an enrichment facility) would greatly improve the inspector function. This type of system can reduce the risk of false or misreported cylinder tare weights, diversion of nuclear material, concealment of excess production, utilization of undeclared cylinders, and misrepresentation of the cylinders contents. This paper will describe a proof-of-concept approach that was designed to evaluate the feasibility of using radio frequency (RF)-based technologies to track individual UF{sub 6} cylinders throughout a portion of their life cycle, and thus demonstrate the potential for improved domestic accountability of materials, and a more effective and efficient method for application of site-level IAEA safeguards. The evaluation system incorporates RF-based identification devices (RFID) which provide a foundation for establishing a reliable, automated, and near real-time tracking system that can be set up to utilize site-specific, rules-based detection algorithms. This paper will report results from a proof-of-concept demonstration at a real enrichment facility that is specifically designed to evaluate both the feasibility of using RF to track cylinders and the durability of the RF equipment to survive the rigors of operational processing and handling. The paper also discusses methods for securely attaching RF devices and describes how the technology can effectively be layered with other safeguard systems and approaches to build a robust system for detecting cylinder diversion. Additionally

  4. SEPARATION OF URANIUM FROM THORIUM

    DOE Patents [OSTI]

    Hellman, N.N.

    1959-07-01

    A process is presented for separating uranium from thorium wherein the ratio of thorium to uranium is between 100 to 10,000. According to the invention the thoriumuranium mixture is dissolved in nitric acid, and the solution is prepared so as to obtain the desired concentration within a critical range of from 4 to 8 N with regard to the total nitrate due to thorium nitrate, with or without nitric acid or any nitrate salting out agent. The solution is then contacted with an ether, such as diethyl ether, whereby uranium is extracted into ihe organic phase while thorium remains in the aqueous phase.

  5. URANIUM RECOVERY FROM NUCLEAR FUEL

    DOE Patents [OSTI]

    Vogel, R.C.; Rodger, W.A.

    1962-04-24

    A process of recovering uranium from a UF/sub 4/-NaFZrF/sub 4/ mixture by spraying the molten mixture at about 200 deg C in nitrogen of super- atmospheric pressure into droplets not larger than 100 microns, and contacting the molten droplets with fluorine at about 200 deg C for 0.01 to 10 seconds in a container the walls of which have a temperature below the melting point of the mixture is described. Uranium hexafluoride is formed and volatilized and the uranium-free salt is solidified. (AEC)

  6. Excess Uranium Inventory Management Plan | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Excess Uranium Inventory Management Plan Excess Uranium Inventory Management Plan The 2013 Excess Uranium Inventory Management Plan describes a framework for the effective...

  7. FLUX COMPOSITION AND METHOD FOR TREATING URANIUM-CONTAINING METAL

    DOE Patents [OSTI]

    Foote, F.

    1958-08-26

    A flux composition is preseated for use with molten uranium and uranium alloys. It consists of about 60% calcium fluoride, 30% calcium chloride and 10% uranium tetrafluoride.

  8. Uranium Processing Facility | Y-12 National Security Complex

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Y-12 Uranium Processing Facility Uranium Processing Facility UPF will be a state-of-the-art, consolidated facility for enriched uranium operations including assembly,...

  9. DOE Uranium Leasing Program 2015 Mitigation Action Plan Activity...

    Energy Savers [EERE]

    DOE Uranium Leasing Program 2015 Mitigation Action Plan Activity Summary Report DOE Uranium Leasing Program 2015 Mitigation Action Plan Activity Summary Report DOE Uranium Leasing ...

  10. Researchers use light to create rare uranium molecule

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Rare uranium molecule Researchers use light to create rare uranium molecule Uranium nitride materials show promise as advanced nuclear fuels due to their high density, high ...

  11. PROCESS FOR THE PURIFICATION OF URANIUM

    DOE Patents [OSTI]

    Rosenfeld, S.

    1959-01-20

    A proccss is described for reclaiming uranium values from aqueous solutions containing U, Fe, Ni, Cu, and Cr comprising treating the solution with NH/sub 3/ to precipitate the: U, Fc, and Cr and leaving Cu and Ni in solution as ammonia complex ions. The precipitate is chlorinated with CCl/sub 4/ at an elevated temperature to convert the U, Tc, and Cr into their chlorides. The more volatile FeCl/sub 3/ and CrCl/sub 3/ are separated from the UCl/sub 4/. The process is used when U is treated in a calutron, and composite solutions are produccd which contain dissolved products of stainless steel.

  12. URANIUM PURIFICATION PROCESS

    DOE Patents [OSTI]

    Ruhoff, J.R.; Winters, C.E.

    1957-11-12

    A process is described for the purification of uranyl nitrate by an extraction process. A solution is formed consisting of uranyl nitrate, together with the associated impurities arising from the HNO/sub 3/ leaching of the ore, in an organic solvent such as ether. If this were back extracted with water to remove the impurities, large quantities of uranyl nitrate will also be extracted and lost. To prevent this, the impure organic solution is extracted with small amounts of saturated aqueous solutions of uranyl nitrate thereby effectively accomplishing the removal of impurities while not allowing any further extraction of the uranyl nitrate from the organic solvent. After the impurities have been removed, the uranium values are extracted with large quantities of water.

  13. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    3. Deliveries of uranium feed by owners and operators of U.S. civilian nuclear power reactors by enrichment country and delivery year, 2013-15 thousand pounds U3O8 equivalent Feed deliveries in 2013 Feed deliveries in 2014 Feed deliveries in 2015 Enrichment country U.S.-origin Foreign-origin Total U.S.-origin Foreign-origin Total U.S.-origin Foreign-origin Total China 0 W W W W W 0 W W France 0 1,606 1,606 0 3,055 3,055 W W 3,299 Germany W W W W W 2,140 W W W Netherlands 1,058 2,773 3,831 0

  14. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    9. Foreign purchases of uranium by U.S. suppliers and owners and operators of U.S. civilian nuclear power reactors by delivery year, 2011-15 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries 2011 2012 2013 2014 2015 U.S. suppliers Foreign purchases 19,318 20,196 23,233 24,199 27,233 Weighted-average price 48.80 46.80 43.25 39.13 40.68 Owners and operators of U.S. civilian nuclear power reactors Foreign purchases 35,071 36,037 34,095 34,404 36,912 Weighted-average

  15. Uranium hexafluoride bibliography

    SciTech Connect (OSTI)

    Burnham, S.L.

    1988-01-01

    This bibliography is a compilation of reports written about the transportation, handling, safety, and processing of uranium hexafluoride. An on-line literature search was executed using the DOE Energy files and the Nuclear Science Abstracts file to identify pertinent reports. The DOE Energy files contain unclassified information that is processed at the Office of Scientific and Technical Information of the US Department of Energy. The reports selected from these files were published between 1974 and 1983. Nuclear Science Abstracts contains unclassified international nuclear science and technology literature published from 1948 to 1976. In addition, scientific and technical reports published by the US Atomic Energy Commission and the US Energy Research and Development Administration, as well as those published by other agencies, universities, and industrial and research organizations, are included in the Nuclear Science Abstracts file. An alphabetical listing of the acronyms used to denote the corporate sponsors follows the bibliography.

  16. PROCESS FOR THE CONCENTRATION OF ORES CONTAINING GOLD AND URANIUM

    DOE Patents [OSTI]

    Gaudin, A.M.; Dasher, J.

    1958-06-10

    ABS>A process is described for concentrating certain low grade uranium and gold bearing ores, in which the gangue is mainly quartz. The production of the concentrate is accomplished by subjecting the crushed ore to a froth floatation process using a fatty acid as a collector in conjunction with a potassium amyl xanthate collector. Pine oil is used as the frothing agent.

  17. Power production and ADS

    SciTech Connect (OSTI)

    Raja, Rajendran; /Fermilab

    2010-03-01

    We describe the power production process in Accelerator Driven Sub-critical systems employing Thorium-232 and Uranium-238 as fuel and examine the demands on the power of the accelerator required.

  18. Domestic Material Content in Molten-Salt Concentrating Solar Power Plants

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Domestic Health Studies and Activities Domestic Health Studies and Activities Purpose The Atomic Energy Act of 1957 - Section 8(a) requires research and development activities relating to the protection of health during research and production activities. The requirement is fulfilled by conducting and supporting health studies and other research activities to determine if DOE workers and people living in communities near DOE sites are adversely affected by exposures to hazardous materials from

  19. Method for cleaning bomb-reduced uranium derbies

    DOE Patents [OSTI]

    Banker, J.G.; Wigginton, H.L.; Beck, D.E.; Holcombe, C.E.

    The concentration of carbon in uranium metal ingots induction cast from derbies prepared by the bomb-reduction of uranium tetrafluoride in the presence of magnesium is effectively reduced to less than 100 ppM by removing residual magnesium fluoride from the surface of the derbies prior to casting. This magnesium fluoride is removed from the derbies by immersing them in an alkali metal salt bath which reacts with and decomposes the magnesium fluoride. A water quenching operation followed by a warm nitric acid bath and a water rinse removes the residual salt and reaction products from the derbies.

  20. Method for cleaning bomb-reduced uranium derbies

    DOE Patents [OSTI]

    Banker, John G.; Wigginton, Hubert L.; Beck, David E.; Holcombe, Cressie E.

    1981-01-01

    The concentration of carbon in uranium metal ingots induction cast from derbies prepared by the bomb-reduction of uranium tetrafluoride in the presence of magnesium is effectively reduced to less than 100 ppm by removing residual magnesium fluoride from the surface of the derbies prior to casting. This magnesium fluoride is removed from the derbies by immersing them in an alkali metal salt bath which reacts with and decomposes the magnesium fluoride. A water quenching operation followed by a warm nitric acid bath and a water rinse removes the residual salt and reaction products from the derbies.

  1. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    The natural UF 6 and enriched UF 6 weighted-average price represent only the U 3 O 8 equivalent uranium-component price specified in the contract for each delivery of natural UF 6 ...

  2. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Deliveries 2011 2012 2013 2014 2015 Purchases 1,668 1,194 W 410 2,702 Weighted-average price ...

  3. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    7 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Table S3b. Weighted-average price of foreign purchases and foreign sales by U.S. ...

  4. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    b. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by purchaser, 2013-15 deliveries" "thousand pounds U3O8 ...

  5. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    6a. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by quantity, 2013-15 deliveries" "thousand pounds U3O8 ...

  6. 2015 Uranium Marketing Annual Survey

    U.S. Energy Information Administration (EIA) Indexed Site

    and enriched UF6 weighted-average price represent only the U3O8 equivalent uranium-component price specified in the contract for each delivery of natural UF6 and enriched UF6, ...

  7. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    2013-15" 2013,2014,2015 "American Fuel Resources, LLC","Advance Uranium Asset Management Ltd.","AREVA AREVA NC, Inc." "AREVA NC, Inc.","AREVA AREVA NC, Inc.","ARMZ ...

  8. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Next Release Date: May 2017 2013 2014 2015 American Fuel Resources, LLC Advance Uranium Asset Management Ltd. AREVA AREVA NC, Inc. AREVA NC, Inc. AREVA AREVA NC, Inc. ARMZ ...

  9. 2015 Uranium Marketing Annual Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Uranium purchased by owners and operators of U.S. civilian nuclear power reactors, ... Purchased from other owners and operators of U.S. civilian nuclear power reactors, other ...

  10. 2015 Uranium Marketing Annual Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    owners and operators of U.S. civilian nuclear power reactors, 1994-2015 Year Feed deliveries by owners and operators of U.S. civilian nuclear power reactors Uranium in fuel ...

  11. Y-12 and uranium history

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    did happen six days after he was given the assignment. The history of uranium at Y-12 began with that decision, which will be commemorated on September 19, 2012, at...

  12. H. R. 5916: A Bill to require the President of the United States to use the Strategic Petroleum Reserve in the event of a domestic energy supply shortage, to amend the Energy Policy and Conservation Act and the Export Administration Act of 1979 to prohibit the exportation of refined petroleum products except under certain circumstances, and for other purposes, introduced in the House of Representatives, One Hundred First Congress, Second Session, October 24, 1990

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    The bill amends the Energy Policy and Conservation Act by making mandatory the use of the Strategic Petroleum Reserve in the event of a domestic energy supply shortage. The restriction on the export of refined petroleum products refers to gasoline, kerosene, heating oils, jet fuel, diesel fuel, residual fuel oil, propane, butane, and any natural liquid or natural gas liquid product refined within the US or entered for consumption within the US. The bill also describes the appointment of special investigator to investigate possible gouging and market manipulation by oil companies and the sense of Congress concerning the cost of deployment and maintenance of United States troops in Saudi Arabia.

  13. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, B.A.

    1983-06-10

    A method is described for measuring the uranium content of aqueous solutions wherein a uranyl phosphate complex is irradiated with a 5 nanosecond pulse of 425 nanometer laser light and resultant 520 nanometer emissions are observed for a period of 50 to 400 microseconds after the pulse. Plotting the natural logarithm of emission intensity as a function of time yields an intercept value which is proportional to uranium concentration.

  14. LIQUID METAL COMPOSITIONS CONTAINING URANIUM

    DOE Patents [OSTI]

    Teitel, R.J.

    1959-04-21

    Liquid metal compositions containing a solid uranium compound dispersed therein is described. Uranium combines with tin to form the intermetallic compound USn/sub 3/. It has been found that this compound may be incorporated into a liquid bath containing bismuth and lead-bismuth components, if a relatively small percentage of tin is also included in the bath. The composition has a low thermal neutron cross section which makes it suitable for use in a liquid metal fueled nuclear reactor.

  15. MELTING AND PURIFICATION OF URANIUM

    DOE Patents [OSTI]

    Spedding, F.H.; Gray, C.F.

    1958-09-16

    A process is described for treating uranium ingots having inner metal portions and an outer oxide skin. The method consists in partially supporting such an ingot on the surface of a grid or pierced plate. A sufficient weight of uranium is provided so that when the mass becomes molten, the oxide skin bursts at the unsupported portions of its bottom surface, allowing molten urantum to flow through the burst skin and into a container provided below.

  16. SURFACE TREATMENT OF METALLIC URANIUM

    DOE Patents [OSTI]

    Gray, A.G.; Schweikher, E.W.

    1958-05-27

    The treatment of metallic uranium to provide a surface to which adherent electroplates can be applied is described. Metallic uranium is subjected to an etchant treatment in aqueous concentrated hydrochloric acid, and the etched metal is then treated to dissolve the resulting black oxide and/or chloride film without destroying the etched metal surface. The oxide or chloride removal is effected by means of moderately concentrated nitric acid in 3 to 20 seconds.

  17. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, Bruce A.

    1986-01-01

    A method is described for measuring the uranium content of aqueous solutions wherein a uranyl phosphate complex is irradiated with a 5 nanosecond pulse of 425 nanometer laser light and resultant 520 nanometer emissions are observed for a period of 50 to 400 microseconds after the pulse. Plotting the natural logarithm of emission intensity as a function of time yields an intercept value which is proportional to uranium concentration.

  18. Uranium Elemental and Isotopic Constraints on Groundwater Flow Beneath the Nopal I Uranium Deposit, Pena Blanca, Mexico

    SciTech Connect (OSTI)

    S.J. Goldstein; M.T. Murrell; A.M. Simmons

    2005-07-11

    The Nopal I uranium deposit in Chihuahua, Mexico, is an excellent analogue for evaluating the fate of spent fuel, associated actinides, and fission products over long time scales for the proposed Yucca Mountain high-level nuclear waste repository. In 2003, three groundwater wells were drilled directly adjacent to (PB-1) and 50 m on either side of the uranium deposit (PB-2 and PB-3) in order to evaluate uranium-series transport in three dimensions. After drilling, uranium concentrations were elevated in all of the three wells (0.1-18 ppm) due to drilling activities and subsequently decreased to {approx}5-20% of initial values over the next several months. The {sup 234}U/{sup 238}U activity ratios were similar for PB-1 and PB-2 (1.005 to 1.079) but distinct for PB-3 (1.36 to 1.83) over this time period, suggesting limited mixing between groundwater from these wells over these short time and length scales. Regional groundwater wells located up to several km from the deposit also have distinct uranium isotopic characteristics and constrain mixing over larger length and time scales. We model the decreasing uranium concentrations in the newly drilled wells with a simple one-dimensional advection-dispersion model, assuming uranium is introduced as a slug to each of the wells and transported as a conservative tracer. Using this model for our data, the relative uranium concentrations are dependent on both the longitudinal dispersion as well as the mean groundwater flow velocity. These parameters have been found to be correlated in both laboratory and field studies of groundwater velocity and dispersion (Klotz et al., 1980). Using typical relationships between velocity and dispersion for field and laboratory studies along with the relationship observed from our uranium data, both velocity (1-10 n/yr) and dispersion coefficient (1E-5 to 1E-2 cm{sup 2}/s) can be derived from the modeling. As discussed above, these relatively small flow velocities and dispersivities agree with

  19. Study on Shielding Requirements for Radioactive Waste Transportation in a Mo-99 Production Plant - 13382

    SciTech Connect (OSTI)

    Melo Rego, Maria Eugenia de; Kazumi Sakata, Solange; Vicente, Roberto; Hiromoto, Goro

    2013-07-01

    Brazil is currently planning to produce {sup 99}Mo from fission of low enriched uranium (LEU) targets. The planned end of irradiation activity of {sup 99}Mo is about 185 TBq (5 kCi) per week to meet the present domestic demand of {sup 99m}Tc generators. The radioactive wastes from the production plant will be transferred to a waste treatment facility at the same site. The total activity of the actinides, fission and activation products present in the wastes can be predicted based on the yields of fission and activation data for the irradiation conditions, such as composition and mass of uranium targets, irradiation time, neutron flux, production schedule, etc., which were in principle already established by the project management. The transportation of the wastes from the production plant to the treatment facility will be done by means of special shielded packages. An assessment of the shielding required for the packages has been done and the results are presented here, aiming at contributing to the design of the waste management facility for the {sup 99}Mo production plant. (authors)

  20. Uranium chloride extraction of transuranium elements from LWR fuel

    DOE Patents [OSTI]

    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.

  1. Uranium chloride extraction of transuranium elements from LWR fuel

    DOE Patents [OSTI]

    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.

  2. Report to the President on Capturing Domestic Competitive Advantage...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Report to the President on Capturing Domestic Competitive Advantage in Advanced Manufacturing Report to the President on Capturing Domestic Competitive Advantage in Advanced ...

  3. Table 22. Domestic Crude Oil First Purchase Prices for Selected...

    U.S. Energy Information Administration (EIA) Indexed Site

    Form EIA-182, "Domestic Crude Oil First Purchase Report." 22. Domestic Crude Oil First Purchase Prices for Selected Crude Streams 44 Energy Information Administration ...

  4. Model Simulating Real Domestic Hot Water Use - Building America...

    Energy Savers [EERE]

    Model Simulating Real Domestic Hot Water Use - Building America Top Innovation Model Simulating Real Domestic Hot Water Use - Building America Top Innovation Image of a pipe ...

  5. Opportunities and Domestic Barriers to Clean Energy Investment...

    Open Energy Info (EERE)

    and Domestic Barriers to Clean Energy Investment in Chile Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Opportunities and Domestic Barriers to Clean Energy Investment...

  6. METHOD OF APPLYING NICKEL COATINGS ON URANIUM

    DOE Patents [OSTI]

    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.

  7. SOLVENT EXTRACTION PROCESS FOR URANIUM RECOVERY

    DOE Patents [OSTI]

    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.

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

    DOE Patents [OSTI]

    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.

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

    DOE Patents [OSTI]

    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.

  10. Compact reaction cell for homogenizing and down-blanding highly enriched uranium metal

    DOE Patents [OSTI]

    McLean, II, William; Miller, Philip E.; Horton, James A.

    1995-01-01

    The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gasses into the reaction chamber, the upper port allowing for the exit of gasses from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gasses into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell.

  11. Compact reaction cell for homogenizing and down-blending highly enriched uranium metal

    DOE Patents [OSTI]

    McLean, W. II; Miller, P.E.; Horton, J.A.

    1995-05-02

    The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gases into the reaction chamber, the upper port allowing for the exit of gases from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gases into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell. 4 figs.

  12. Nuclear forensic analysis of uranium oxide powders interdicted in Victoria, Australia

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael; Williams, Ross; Lindvall, Rachel; Eppich, Gary; Roberts, Sarah; Borg, Lars; Gaffney, Amy; Plaue, Jonathan; et al

    2015-04-13

    Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (~1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K2(UO2)3O4·4H2O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (~380 μg/g total elemental impurities). The chemical form of the uranium was primarily UO3·2H2O, with minor phases ofmore » U3O8 and UO2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.« less

  13. Uranium reference materials

    SciTech Connect (OSTI)

    Donivan, S.; Chessmore, R.

    1987-07-01

    The Technical Measurements Center has prepared uranium mill tailings reference materials for use by remedial action contractors and cognizant federal and state agencies. Four materials were prepared with varying concentrations of radionuclides, using three tailings materials and a river-bottom soil diluent. All materials were ground, dried, and blended thoroughly to ensure homogeneity. The analyses on which the recommended values for nuclides in the reference materials are based were performed, using independent methods, by the UNC Geotech (UNC) Chemistry Laboratory, Grand Junction, Colorado, and by C.W. Sill (Sill), Idaho National Engineering Laboratory, Idaho Falls, Idaho. Several statistical tests were performed on the analytical data to characterize the reference materials. Results of these tests reveal that the four reference materials are homogeneous and that no large systematic bias exists between the analytical methods used by Sill and those used by TMC. The average values for radionuclides of the two data sets, representing an unbiased estimate, were used as the recommended values for concentrations of nuclides in the reference materials. The recommended concentrations of radionuclides in the four reference materials are provided. Use of these reference materials will aid in providing uniform standardization among measurements made by remedial action contractors. 11 refs., 9 tabs.

  14. The strategy on rehabilitation of the former uranium facilities at the 'Pridneprovsky chemical plant' in Ukraine

    SciTech Connect (OSTI)

    Voitsekhovich, O.; Lavrova, T. [Ukrainian Hydrometeorological Institute, Kiev (Ukraine); Skalskiy, A.S. [Institute of Geological Sciences of Ac.of Sc., Kiev (Ukraine); Ryazantsev, V.F. [State Nuclear Regulatory Committee of Ukraine, 9/11 Arsenalna str., Kyiv-11, 01011 (Ukraine)

    2007-07-01

    This paper describes current status of the former Uranium Facilities at the Pridneprovsky Chemical Plant in Ukraine, which are currently under development of action plan for its territory rehabilitation. The monitoring data carried out during recent several years show its impact to the Environment and gives a basis for justification of the number of measures aiming to reduce radiological and ecological risks of the Uranium tailings situated at the territory of PChP. The monitoring data and strategy for its remediation are considered in the presentation. Uranium mining has been intensively conducted in Ukraine since the end of the 40-s. Most of the uranium deposits have been explored in the Dnieper river basin, while some smaller deposits can be found within the basins of the Southern Bug and Severskiy Donets rivers. There also several large Uranium Milling facilities were in operation since the end of the 40-s till 1991, when due to disintegration of the former Soviet Union system the own uranium production has been significantly declined. The Milling Plant and Uranium extraction Facilities in ZhevtiVody is still in operation with UkrAtomprom Industrial Consortium. Therefore rehabilitation programme for all Uranium facilities in this site are in duty of the East Mining Combine and the Consortium. The most difficult case is to provide rehabilitation Action Plan for Uranium tailings and number of other facilities situated in Dnieprodzerzhinsk town and which were in operation by the former State Industrial Enterprise Pridneprovskiy Chemical Plant (PChP). In past PChP was one of the largest Uranium Milling facilities of the Former Soviet Union and has been in operation since 1948 till 1991. During Soviet time the Uranium extraction at this legacy site has been carried out using the ore raw products delivered also from Central Asia, Germany and Checz Republic. After extraction the uranium residue has been putting to the nearest landscape depressions at the vicinity of

  15. Small cell experiments for electrolytic reduction of uranium oxides to uranium metal using fluoride salts

    SciTech Connect (OSTI)

    Haas, P.A.; Adcock, P.W.; Coroneos, A.C.; Hendrix, D.E. )

    1994-08-01

    Electrolytic reduction of uranium oxide was proposed for the preparation of uranium metal feed for the atomic vapor laser isotope separation (AVLIS) process. A laboratory cell of 25-cm ID was operated to obtain additional information in areas important to design and operation of a pilot plant cell. Reproducible test results and useful operating and control procedures were demonstrated. About 20 kg of uranium metal of acceptable purity were prepared. A good supply of dissolved UO[sub 2] feed at the anode is the most important controlling requirement for efficient cell operation. A large fraction of the cell current is nonproductive in that it does not produce a metal product nor consume carbon anodes. All useful test conditions gave some reduction of UF[sub 4] to produce CF[sub 4] in addition to the reduction of UO[sub 2], but the fraction of metal from the reduction of UF[sub 4] can be decreased by increasing the concentration of dissolved UO[sub 2]. Operation of large continuous cells would probably be limited to current efficiencies of less than 60 pct, and more than 20 pct of the metal would result from the reduction of UF[sub 4].

  16. Uranium Leasing Program Environmental Documents | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Environmental Documents Uranium Leasing Program Environmental Documents Uranium Leasing Program 2015 Mitigation Action Plan Activity Summary Report (March 2016) The DOE Uranium Leasing Program's 2015 Mitigation Action Plan Activity Summary fulfills the mitigation plan's requirement to annually notify the public of mitigation activities completed by Uranium Leasing Program lessees. Uranium Leasing Program Mitigation Action Plan for the Final Uranium Leasing Program Programmatic Environmental

  17. PROCESSES FOR SEPARATING AND RECOVERING CONSTITUENTS OF NEUTRON IRRADIATED URANIUM

    DOE Patents [OSTI]

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

    1959-11-10

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

  18. Performance Assessment Transport Modeling of Uranium at the Area 5 Radioactive Waste Management Site at the Nevada National Security Site

    SciTech Connect (OSTI)

    NSTec Radioactive Waste

    2010-10-12

    Following is a brief summary of the assumptions that are pertinent to the radioactive isotope transport in the GoldSim Performance Assessment model of the Area 5 Radioactive Waste Management Site, with special emphasis on the water-phase reactive transport of uranium, which includes depleted uranium products.

  19. Uranium Biomineralization By Natural Microbial Phosphatase Activities...

    Office of Scientific and Technical Information (OSTI)

    Finally, the minerals produced during this process are stable in low pH conditions or ... strategy to uranium bioreduction in low pH uranium-contaminated environments. ...

  20. Absorption of Thermal Neutrons in Uranium

    DOE R&D Accomplishments [OSTI]

    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.

  1. Uranium Marketing Annual Report - Energy Information Administration

    Gasoline and Diesel Fuel Update (EIA)

    Uranium purchases and prices Owners and operators of U.S. civilian nuclear power reactors ... Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors during 2015 ...

  2. Inherently safe in situ uranium recovery

    SciTech Connect (OSTI)

    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.

  3. RECOVERY OF URANIUM VALUES FROM URANIUM BEARING RAW MATERIALS

    DOE Patents [OSTI]

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

    1959-06-16

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

  4. PROCESS FOR THE RECOVERY OF URANIUM

    DOE Patents [OSTI]

    Morris, G.O.

    1955-06-21

    This patent relates to a process for the recovery of uranium from impure uranium tetrafluoride. The process consists essentially of the steps of dissolving the impure uranium tetrafluoride in excess dilute sulfuric acid in the presence of excess hydrogen peroxide, precipitating ammonium uranate from the solution so formed by adding an excess of aqueous ammonia, dissolving the precipitate in sulfuric acid and adding hydrogen peroxide to precipitate uranium peroxdde.

  5. METHOD OF APPLYING COPPER COATINGS TO URANIUM

    DOE Patents [OSTI]

    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.

  6. Uranium Management and Policy | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Uranium Management and Policy Uranium Management and Policy The Paducah Gaseous Diffusion Plant is located 3 miles south of the Ohio River and is 12 miles west of Paducah, Kentucky. Paducah remains the only operating gaseous diffusion uranium enrichment plant in the United States. The Paducah Gaseous Diffusion Plant is located 3 miles south of the Ohio River and is 12 miles west of Paducah, Kentucky. Paducah remains the only operating gaseous diffusion uranium enrichment plant in the United

  7. URANIUM BISMUTHIDE DISPERSION IN MOLTEN METAL

    DOE Patents [OSTI]

    Teitel, R.J.

    1959-10-27

    The formation of intermetallic bismuth compounds of thorium or uranium dispersed in a liquid media containing bismuth and lead is described. A bismuthide of uranium dispersed in a liquid metal medium is formed by dissolving uranium in composition of lead and bismuth containing less than 80% lead and lowering the temperature of the composition to a temperature below the point at which the solubility of uranium is exceeded and above the melting point of the composition.

  8. Uranium Downblending and Disposition Project Technology Readiness

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Assessment | Department of Energy Uranium Downblending and Disposition Project Technology Readiness Assessment Uranium Downblending and Disposition Project Technology Readiness Assessment Full Document and Summary Versions are available for download Uranium Downblending and Disposition Project Technology Readiness Assessment (1.11 MB) Summary - Uranium233 Downblending and Disposition Project (146.5 KB) More Documents & Publications Compilation of TRA Summaries EA-1574: Final

  9. The directory of United States coal & technology export resources. Profiles of domestic US corporations, associations and public entities, nationwide, which offer products or services suitable for export, relating to coal and its utilization

    SciTech Connect (OSTI)

    Not Available

    1994-01-01

    The purpose of this directory is to provide a listing of available U.S. coal and coal related resources to potential purchasers of those resources abroad. The directory lists business entities within the US which offer coal related resources, products and services for sale on the international market. Each listing is intended to describe the particular business niche or range of product and/or services offered by a particular company. The listing provides addresses, telephones, and telex/fax for key staff in each company committed to the facilitation of international trade. The content of each listing has been formulated especially for this directory and reflects data current as of the date of this edition. The directory listings are divided into four primary classifications: coal resources; technology resources; support services; and financing and resource packaging. The first three of which are subdivided as follows: Coal Resources -- coal derivatives, coal exporters, and coal mining; Technology Resources -- advanced utilization, architects and engineers, boiler equipment, emissions control and waste disposal systems, facility construction, mining equipment, power generation systems, technical publications, and transport equipment; Support Services -- coal transport, facility operations, freight forwarders, sampling services and equipment, and technical consultants. Listings for the directory were solicited on the basis of this industry breakdown. Each of the four sections of this directory begins with a matrix illustrating which companies fall within the particular subclassifications specific to that main classification. A general alphabetical index of companies and an index by product/service classification are provided following the last section of the directory.

  10. Uranium distribution and geology in the Fish Lake surficial uranium deposit, Esmeralda County, Nevada

    SciTech Connect (OSTI)

    Macke, D.L.; Schumann, R.R.; Otton, J.K.

    1990-01-01

    This paper reports on approximately 675 acres of uranium-enriched lacustrine and marsh sediments in Fish Lake Valley, in southern Nevada and California. Uranium concentrations from 253 samples averaged 64.3 ppm uranium, with a range from 6 to 800 ppm. Uranium was supplied to the marsh sediments by ground water derived from Tertiary volcanic rocks of the Silver Peak Range. Reconnaissance sampling in the surrounding areas shows minor enrichment of uranium in other wetland areas.

  11. CATALYZED OXIDATION OF URANIUM IN CARBONATE SOLUTIONS

    DOE Patents [OSTI]

    Clifford, W.E.

    1962-05-29

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

  12. High strength uranium-tungsten alloy process

    DOE Patents [OSTI]

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

    1990-01-01

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

  13. High strength uranium-tungsten alloys

    DOE Patents [OSTI]

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

    1991-01-01

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

  14. final ERI-2142 18-1501 Analysis of Potential Effects on Domestic...

    Office of Environmental Management (EM)

    ... may contain natural uranium pellets, such that some uranium does not require enrichment; the feed material for these natural uranium pellets may also be made from UO 2 , ...

  15. METHOD OF PROTECTIVELY COATING URANIUM

    DOE Patents [OSTI]

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

    1959-02-01

    A method is described for protectively coating uranium with zine comprising cleaning the U for coating by pickling in concentrated HNO/sub 3/, dipping the cleaned U into a bath of molten zinc between 430 to 600 C and containing less than 0 01% each of Fe and Pb, and withdrawing and cooling to solidify the coating. The zinccoated uranium may be given a; econd coating with another metal niore resistant to the corrosive influences particularly concerned. A coating of Pb containing small proportions of Ag or Sn, or Al containing small proportions of Si may be applied over the zinc coatings by dipping in molten baths of these metals.

  16. Uranium-233 purification and conversion to stabilized ceramic grade urania for LWBR fuel fabrication (LWBR Development Program)

    SciTech Connect (OSTI)

    Lloyd, R.

    1980-10-01

    High purity ceramic grade urania (/sup 233/UO/sub 2/) used in manufacturing the fuel for the Light Water Breeder Reactor (LWBR) core was made from uranium-233 that was obtained by irradiating thoria under special conditions to result in not more than 10 ppM of uranium-232 in the recovered uranium-233 product. A developmental study established the operating parameters of the conversion process for transforming the uranium-233 into urania powder with the appropriate chemical and physical attributes for use in fabricating the LWBR core fuel. This developmental study included the following: (a) design of an ion exchange purification process for removing the gamma-emitting alpha-decay daughters of uranium-232, to reduce the gamma-radiation field of the uranium-233 during LWBR fuel manufacture; (b) definition of the parameters for precipitating the uranium-233 as ammonium uranate (ADU) and for reducing the ADU with hydrogen to yield a urania conversion product of the proper particle size, surface area and sinterability for use in manufacturing the LWBR fuel; (c) establishment of parameters and design of equipment for stabilizing the urania conversion product to prevent it from undergoing excessive oxidation on exposure to the air during LWBR fuel manufacturing operations; and (d) development of a procedure and a facility to reprocess the unirradiated thoria-urania fuel scrap from the LWBR core manufacturing operations to recover the uranium-233 and convert it into high purity ceramic grade urania for LWBR core fabrication.

  17. Electron Backscatter Diffraction (EBSD) Characterization of Uranium and Uranium Alloys

    SciTech Connect (OSTI)

    McCabe, Rodney J.; Kelly, Ann Marie; Clarke, Amy J.; Field, Robert D.; Wenk, H. R.

    2012-07-25

    Electron backscatter diffraction (EBSD) was used to examine the microstructures of unalloyed uranium, U-6Nb, U-10Mo, and U-0.75Ti. For unalloyed uranium, we used EBSD to examine the effects of various processes on microstructures including casting, rolling and forming, recrystallization, welding, and quasi-static and shock deformation. For U-6Nb we used EBSD to examine the microstructural evolution during shape memory loading. EBSD was used to study chemical homogenization in U-10Mo, and for U-0.75Ti, we used EBSD to study the microstructure and texture evolution during thermal cycling and deformation. The studied uranium alloys have significant microstructural and chemical differences and each of these alloys presents unique preparation challenges. Each of the alloys is prepared by a sequence of mechanical grinding and polishing followed by electropolishing with subtle differences between the alloys. U-6Nb and U-0.75Ti both have martensitic microstructures and both require special care in order to avoid mechanical polishing artifacts. Unalloyed uranium has a tendency to rapidly oxidize when exposed to air and a two-step electropolish is employed, the first step to remove the damaged surface layer resulting from the mechanical preparation and the second step to passivate the surface. All of the alloying additions provide a level of surface passivation and different one and two step electropolishes are employed to create good EBSD surfaces. Because of its low symmetry crystal structure, uranium exhibits complex deformation behavior including operation of multiple deformation twinning modes. EBSD was used to observe and quantify twinning contributions to deformation and to examine the fracture behavior. Figure 1 shows a cross section of two mating fracture surfaces in cast uranium showing the propensity of deformation twinning and intergranular fracture largely between dissimilarly oriented grains. Deformation of U-6Nb in the shape memory regime occurs by the motion

  18. Advanced Manufacturing Initiative Improves Turbine Blade Productivity...

    Broader source: Energy.gov (indexed) [DOE]

    and create U.S. jobs by improving labor productivity in wind turbine blade construction. ... Certain components of wind turbine blades are naturally more suitable to domestic ...

  19. Uranium Processing Facility Project Celebrates Changing the Skyline of Y-12

    National Nuclear Security Administration (NNSA)

    | National Nuclear Security Administration | (NNSA) fieldoffices / Welcome to the NNSA Production Office / NPO News Releases Uranium Processing Facility Project Celebrates Changing the Skyline of Y-12 August 25, 2016 Groundbreaking Ceremony Held for the Construction Support Building OAK RIDGE, Tenn.-The Uranium Processing Facility (UPF) Project celebrated the groundbreaking for its Construction Support Building (CSB) today, signifying the first building construction activity where the

  20. Production

    Broader source: Energy.gov [DOE]

    Algae production R&D focuses on exploring resource use and availability, algal biomass development and improvements, characterizing algal biomass components, and the ecology and engineering of cultivation systems.