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We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


1

Packaging and Disposal of a Radium-beryllium Source using Depleted Uranium Polyethylene Composite Shielding  

SciTech Connect (OSTI)

Two, 111-GBq (3 Curie) radium-beryllium (RaBe) sources were in underground storage at the Brookhaven National Laboratory (BNL) since 1988. These sources originated from the Princeton Plasma Physics Laboratory (PPPL) where they were used to calibrate neutron detection diagnostics. In 1999, PPPL and BNL began a collaborative effort to expand the use of an innovative pilot-scale technology and bring it to full-scale deployment to shield these sources for eventual transport and burial at the Hanford Burial site. The transport/disposal container was constructed of depleted uranium oxide encapsulated in polyethylene to provide suitable shielding for both gamma and neutron radiation. This new material can be produced from recycled waste products (depleted uranium and polyethylene), is inexpensive, and can be disposed with the waste, unlike conventional lead containers, thus reducing exposure time for workers. This paper will provide calculations and information that led to the initial design of the shielding. We will also describe the production-scale processing of the container, cost, schedule, logistics, and many unforeseen challenges that eventually resulted in the successful fabrication and deployment of this shield. We will conclude with a description of the final configuration of the shielding container and shipping package along with recommendations for future shielding designs.

Keith Rule; Paul Kalb; Pete Kwaschyn

2003-02-11T23:59:59.000Z

2

Use of depleted uranium metal as cask shielding in high-level waste storage, transport, and disposal systems  

SciTech Connect (OSTI)

The US DOE has amassed over 555,000 metric tons of depleted uranium from its uranium enrichment operations. Rather than dispose of this depleted uranium as waste, this study explores a beneficial use of depleted uranium as metal shielding in casks designed to contain canisters of vitrified high-level waste. Two high-level waste storage, transport, and disposal shielded cask systems are analyzed. The first system employs a shielded storage and disposal cask having a separate reusable transportation overpack. The second system employs a shielded combined storage, transport, and disposal cask. Conceptual cask designs that hold 1, 3, 4 and 7 high-level waste canisters are described for both systems. In all cases, cask design feasibility was established and analyses indicate that these casks meet applicable thermal, structural, shielding, and contact-handled requirements. Depleted uranium metal casting, fabrication, environmental, and radiation compatibility considerations are discussed and found to pose no serious implementation problems. About one-fourth of the depleted uranium inventory would be used to produce the casks required to store and dispose of the nearly 15,400 high-level waste canisters that would be produced. This study estimates the total-system cost for the preferred 7-canister storage and disposal configuration having a separate transportation overpack would be $6.3 billion. When credits are taken for depleted uranium disposal cost, a cost that would be avoided if depleted uranium were used as cask shielding material rather than disposed of as waste, total system net costs are between $3.8 billion and $5.5 billion.

Yoshimura, H.R.; Ludwigsen, J.S.; McAllaster, M.E. [and others

1996-09-01T23:59:59.000Z

3

PACKAGING AND DISPOSAL OF A RADIUM BERYLLIUM SOURCE USING DEPLETED URANIUM POLYETHYLENE COMPOSITE SHIELDING.  

SciTech Connect (OSTI)

Two, 111 GBq (3 Curie) radium-beryllium (RaBe) sources were in underground storage at the Brookhaven National Laboratory (BNL) since 1988. These sources originated from Princeton Plasma Physics Laboratory (PPPL) where they were used to calibrate neutron detection diagnostics. In 1999, PPPL and BNL began a collaborative effort to expand the use of an innovative pilot-scale technology and bring it to full-scale deployment to shield these sources for eventual transport and burial at the Hanford Burial site. The transport/disposal container was constructed of depleted uranium oxide encapsulated in polyethylene to provide suitable shielding for both gamma and neutron radiation. This new material can be produced from recycled waste products (DU and polyethylene), is inexpensive, and can be disposed with the waste, unlike conventional lead containers, thus reducing exposure time for workers. This paper will provide calculations and information that led to the initial design of the shielding. We will also describe the production-scale processing of the container, cost, schedule, logistics, and many unforeseen challenges that eventually resulted in the successful fabrication and deployment of this shield. We will conclude with a description of the final configuration of the shielding container and shipping package along with recommendations for future shielding designs.

RULE,K.; KALB,P.; KWASCHYN,P.

2003-02-23T23:59:59.000Z

4

Depleted uranium management alternatives  

SciTech Connect (OSTI)

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.

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

1994-08-01T23:59:59.000Z

5

Depleted Uranium Technical Brief  

E-Print Network [OSTI]

Depleted Uranium Technical Brief United States Environmental Protection Agency Office of Air and Radiation Washington, DC 20460 EPA-402-R-06-011 December 2006 #12;#12;Depleted Uranium Technical Brief EPA of Radiation and Indoor Air Radiation Protection Division ii #12;iii #12;FOREWARD The Depleted Uranium

6

Depleted uranium: A DOE management guide  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) has a management challenge and financial liability in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. The annual storage and maintenance cost is approximately $10 million. This report summarizes several studies undertaken by the DOE Office of Technology Development (OTD) to evaluate options for long-term depleted uranium management. Based on studies conducted to date, the most likely use of the depleted uranium is for shielding of spent nuclear fuel (SNF) or vitrified high-level waste (HLW) containers. The alternative to finding a use for the depleted uranium is disposal as a radioactive waste. Estimated disposal costs, utilizing existing technologies, range between $3.8 and $11.3 billion, depending on factors such as applicability of the Resource Conservation and Recovery Act (RCRA) and the location of the disposal site. The cost of recycling the depleted uranium in a concrete based shielding in SNF/HLW containers, although substantial, is comparable to or less than the cost of disposal. Consequently, the case can be made that if DOE invests in developing depleted uranium shielded containers instead of disposal, a long-term solution to the UF{sub 6} problem is attained at comparable or lower cost than disposal as a waste. Two concepts for depleted uranium storage casks were considered in these studies. The first is based on standard fabrication concepts previously developed for depleted uranium metal. The second converts the UF{sub 6} to an oxide aggregate that is used in concrete to make dry storage casks.

NONE

1995-10-01T23:59:59.000Z

7

Depleted uranium hexafluoride: Waste or resource?  

SciTech Connect (OSTI)

the US Department of Energy is evaluating technologies for the storage, disposal, or re-use of depleted uranium hexafluoride (UF{sub 6}). This paper discusses the following options, and provides a technology assessment for each one: (1) conversion to UO{sub 2} for use as mixed oxide duel, (2) conversion to UO{sub 2} to make DUCRETE for a multi-purpose storage container, (3) conversion to depleted uranium metal for use as shielding, (4) conversion to uranium carbide for use as high-temperature gas-cooled reactor (HTGR) fuel. In addition, conversion to U{sub 3}O{sub 8} as an option for long-term storage is discussed.

Schwertz, N.; Zoller, J.; Rosen, R.; Patton, S. [Lawrence Livermore National Lab., CA (United States); Bradley, C. [USDOE Office of Nuclear Energy, Science, Technology, Washington, DC (United States); Murray, A. [SAIC (United States)

1995-07-01T23:59:59.000Z

8

Depleted uranium disposal options evaluation  

SciTech Connect (OSTI)

The Department of Energy (DOE), Office of Environmental Restoration and Waste Management, has chartered a study to evaluate alternative management strategies for depleted uranium (DU) currently stored throughout the DOE complex. Historically, DU has been maintained as a strategic resource because of uses for DU metal and potential uses for further enrichment or for uranium oxide as breeder reactor blanket fuel. This study has focused on evaluating the disposal options for DU if it were considered a waste. This report is in no way declaring these DU reserves a ``waste,`` but is intended to provide baseline data for comparison with other management options for use of DU. To PICS considered in this report include: Retrievable disposal; permanent disposal; health hazards; radiation toxicity and chemical toxicity.

Hertzler, T.J.; Nishimoto, D.D.; Otis, M.D. [Science Applications International Corp., Idaho Falls, ID (United States). Waste Management Technology Div.

1994-05-01T23:59:59.000Z

9

Review The Toxicity of Depleted Uranium  

E-Print Network [OSTI]

Abstract: Depleted uranium (DU) is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a clear and defined set of symptoms. Chronic low-dose, or subacute, exposure to depleted uranium alters the appearance of milestones in developing organisms. Adult animals that were exposed to depleted uranium during development display persistent alterations in behavior, even after cessation of depleted uranium exposure. Adult animals exposed to depleted uranium demonstrate altered behaviors and a variety of alterations to brain chemistry. Despite its reduced level of radioactivity evidence continues to accumulate that depleted uranium, if ingested, may pose a radiologic hazard. The current state of knowledge concerning DU is discussed.

Wayne Briner

10

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents [OSTI]

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Forsberg, Charles W. (Oak Ridge, TN)

1998-01-01T23:59:59.000Z

11

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents [OSTI]

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Forsberg, C.W.

1998-11-03T23:59:59.000Z

12

Depleted Uranium Hexafluoride (DUF6) Fully Operational at the...  

Office of Environmental Management (EM)

Depleted Uranium Hexafluoride (DUF6) Fully Operational at the Portsmouth and Paducah Gaseous Diffusion Sites Depleted Uranium Hexafluoride (DUF6) Fully Operational at the...

13

Depleted uranium disposition study -- Supplement, Revision 1  

SciTech Connect (OSTI)

The Department of Energy Office of Weapons and Materials Planning has requested a supplemental study to update the recent Depleted Uranium Disposition report. This supplemental study addresses new disposition alternatives and changes in status.

Becker, G.W.

1993-11-01T23:59:59.000Z

14

Molten-Salt Depleted-Uranium Reactor  

E-Print Network [OSTI]

The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results demonstrate how these reactors can possess and realize extraordinary excellent characteristics, no prompt critical, long-term safe and stable operation with negative feedback, closed uranium-plutonium cycle chain within the vessel, normal operation only with depleted-uranium, and depleted-uranium high burnup in reality, to realize with fission nuclear energy sufficiently satisfying humanity long-term energy resource needs, as well as thoroughly solve the challenges of nuclear criticality safety, uranium resource insuffic...

Dong, Bao-Guo; Gu, Ji-Yuan

2015-01-01T23:59:59.000Z

15

The ultimate disposition of depleted uranium  

SciTech Connect (OSTI)

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.

Lemons, T.R. [Uranium Enrichment Organization, Oak Ridge, TN (United States)

1991-12-31T23:59:59.000Z

16

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

DOE Patents [OSTI]

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.

Kalb, Paul D. (Wading River, NY); Adams, Jay W. (Stony Brook, NY); Lageraaen, Paul R. (Seaford, NY); Cooley, Carl R. (Gaithersburg, MD)

2000-02-29T23:59:59.000Z

17

Capstone Depleted Uranium Aerosols: Generation and Characterization  

SciTech Connect (OSTI)

In a study designed to provide an improved scientific basis for assessing possible health effects from inhaling depleted uranium (DU) aerosols, a series of DU penetrators was fired at an Abrams tank and a Bradley fighting vehicle. A robust sampling system was designed to collect aerosols in this difficult environment and continuously monitor the sampler flow rates. Aerosols collected were analyzed for uranium concentration and particle size distribution as a function of time. They were also analyzed for uranium oxide phases, particle morphology, and dissolution in vitro. The resulting data provide input useful in human health risk assessments.

Parkhurst, MaryAnn; Szrom, Fran; Guilmette, Ray; Holmes, Tom; Cheng, Yung-Sung; Kenoyer, Judson L.; Collins, John W.; Sanderson, T. Ellory; Fliszar, Richard W.; Gold, Kenneth; Beckman, John C.; Long, Julie

2004-10-19T23:59:59.000Z

18

Depleted uranium plasma reduction system study  

SciTech Connect (OSTI)

A system life-cycle cost study was conducted of a preliminary design concept for a plasma reduction process for converting depleted uranium to uranium metal and anhydrous HF. The plasma-based process is expected to offer significant economic and environmental advantages over present technology. Depleted Uranium is currently stored in the form of solid UF{sub 6}, of which approximately 575,000 metric tons is stored at three locations in the U.S. The proposed system is preconceptual in nature, but includes all necessary processing equipment and facilities to perform the process. The study has identified total processing cost of approximately $3.00/kg of UF{sub 6} processed. Based on the results of this study, the development of a laboratory-scale system (1 kg/h throughput of UF6) is warranted. Further scaling of the process to pilot scale will be determined after laboratory testing is complete.

Rekemeyer, P.; Feizollahi, F.; Quapp, W.J.; Brown, B.W.

1994-12-01T23:59:59.000Z

19

Depleted Uranium in Kosovo Post-Conflict Environmental Assessment  

E-Print Network [OSTI]

2.1 UNEP’s role in post-conflict environmental assessment................................................9 2.2 Depleted uranium............................................................10

Unep Scientific; Mission Kosovo

20

Assessment of Preferred Depleted Uranium Disposal Forms  

SciTech Connect (OSTI)

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.

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

2000-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

The ultimate disposition of depleted uranium  

SciTech Connect (OSTI)

Significant amounts of the depleted uranium (DU) created by past uranium enrichment activities have been sold, disposed of commercially, or utilized by defense programs. In recent years, however, the demand for DU has become quite small compared to quantities available, and within the US Department of Energy (DOE) there is concern for any risks and/or cost liabilities that might be associated with the ever-growing inventory of this material. As a result, Martin Marietta Energy Systems, Inc. (Energy Systems), was asked to review options and to develop a comprehensive plan for inventory management and the ultimate disposition of DU accumulated at the gaseous diffusion plants (GDPs). An Energy Systems task team, under the chairmanship of T. R. Lemons, was formed in late 1989 to provide advice and guidance for this task. This report reviews options and recommends actions and objectives in the management of working inventories of partially depleted feed (PDF) materials and for the ultimate disposition of fully depleted uranium (FDU). Actions that should be considered are as follows. (1) Inspect UF{sub 6} cylinders on a semiannual basis. (2) Upgrade cylinder maintenance and storage yards. (3) Convert FDU to U{sub 3}O{sub 8} for long-term storage or disposal. This will include provisions for partial recovery of costs to offset those associated with DU inventory management and the ultimate disposal of FDU. Another recommendation is to drop the term tails'' in favor of depleted uranium'' or DU'' because the tails'' label implies that it is waste.'' 13 refs.

Not Available

1990-12-01T23:59:59.000Z

22

Uranio impoverito: perché? (Depleted uranium: why?)  

E-Print Network [OSTI]

In this paper we develop a simple model of the penetration process of a long rod through an uniform target. Applying the momentum and energy conservation laws, we derive an analytical relation which shows how the penetration depth depends upon the density of the rod, given a fixed kinetic energy. This work was sparked off by the necessity of understanding the effectiveness of high density penetrators (e.g. depleted uranium penetrators) as anti-tank weapons.

Germano D'Abramo

2003-06-05T23:59:59.000Z

23

Processing depleted uranium quad alloy penetrator rods  

SciTech Connect (OSTI)

Two depleted uranium (DU) quad alloys were cast, extruded and rolled to produce penetrator rods. The two alloy combinations were (1) 1 wt % molybdenum (Mo), 1 wt % niobium (Nb), and 0.75 wt % titanium (Ti); and (2) 1 wt % tantalum (Ta), 1 wt % Nb, and 0.75 wt % Ti. This report covers the processing and results with limited metallographic information available. The two alloys were each vacuum induction melted (VIM) into an 8-in. log, extruded into a 3-in. log, then cut into 4 logs and extruded at 4 different temperatures into 0.8-in. bars. From the 8 conditions (2 alloys, 4 extrusion temperatures each), 10 to 13 16-in. rods were cut for rolling and swaging. Due to cracking problems, the final processing changed from rolling and swaging to limited rolling and heat treating. The contracted work was completed with the delivery of 88 rods to Dr. Zabielski. 28 figs.

Bokan, S.L.

1987-02-19T23:59:59.000Z

24

DOE Announces Transfer of Depleted Uranium to Advance the U.S...  

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

Transfer of Depleted Uranium to Advance the U.S. National Security Interests, Extend Operations at Paducah Gaseous Diffusion Plant DOE Announces Transfer of Depleted Uranium to...

25

The Hazard Posed by Depleted Uranium Munitions  

E-Print Network [OSTI]

This paper assesses the radiological and chemical hazards resulting from the use of depleted uranium (DU) munitions. Due to the low radioactivity of DU, radiological hazards to individuals would become significant in comparison to natural background radiation doses only in cases of prolonged contact---for example, when shards of a DU penetrator remain embedded in a soldier's body. Although the radiation doses to virtually all civilians would be very low, the cumulative "population dose" resulting from the dispersal of hundreds of tons of DU, as occurred during the Gulf War, could result in up to ten cancer deaths. It is highly unlikely that exposures of persons downwind from the use of DU munitions or consuming food or water contaminated by DU dust would reach the estimated threshold for chemical heavy-metal effects. The exposures of soldiers in vehicles struck by DU munitions could be much higher, however, and persons who subsequently enter such vehicles without adequate respiratory protection could potentially be at risk. Soldiers should be trained to avoid unnecessary exposure to DU, and vehicles struck by DU munitions should be made inaccessible to curious civilians. INTRODUCTION

Steve Fetter And; Steve Fetter A

26

Status Report and Proposal Concerning the Supply of Depleted Uranium Metal Bands for a Particle Detector  

E-Print Network [OSTI]

Status Report and Proposal Concerning the Supply of Depleted Uranium Metal Bands for a Particle Detector

1980-01-01T23:59:59.000Z

27

Conversion of depleted uranium hexafluoride to a solid uranium compound  

DOE Patents [OSTI]

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.

Rothman, Alan B. (Willowbrook, IL); Graczyk, Donald G. (Lemont, IL); Essling, Alice M. (Elmhurst, IL); Horwitz, E. Philip (Naperville, IL)

2001-01-01T23:59:59.000Z

28

Incentives for the use of depleted uranium alloys as transport cask containment structure  

SciTech Connect (OSTI)

Radioactive material transport casks use either lead or depleted uranium (DU) as gamma-ray shielding material. Stainless steel is conventionally used for structural containment. If a DU alloy had sufficient properties to guarantee resistance to failure during both nominal use and accident conditions to serve the dual-role of shielding and containment, the use of other structure materials (i.e., stainless steel) could be reduced. (It is recognized that lead can play no structural role.) Significant reductions in cask weight and dimensions could then be achieved perhaps allowing an increase in payload. The mechanical response of depleted uranium has previously not been included in calculations intended to show that DU-shielded transport casks will maintain their containment function during all conditions. This paper describesa two-part study of depleted uranium alloys: First, the mechanical behavior of DU alloys was determined in order to extend the limited set of mechanical properties reported in the literature. The mechanical properties measured include the tensile behavior the impact energy. Fracture toughness testing was also performed to determine the sensitivity of DU alloys to brittle fracture. Fracture toughness is the inherent material property which quantifies the fracmm resistance of a material. Tensile strength and ductility are significant in terms of other failure modes, however, as win be discussed. These mechanical properties were then input into finite element calculations of cask response to loading conditions to quantify the potential for claiming structural credit for DU. (The term structural credit'' describes whether a material has adequate properties to allow it to assume a positive role in withstanding structural loadings.)

McConnell, P [GRAM, Inc., Albuquerque, NM (United States); Salzbrenner, R; Wellman, G W; Sorenson, K B [Sandia National Labs., Albuquerque, NM (United States)

1992-01-01T23:59:59.000Z

29

EIS-0269: Long-Term Management of Depleted Uranium Hexaflouride  

Broader source: Energy.gov [DOE]

The U.S. Department of Energy (DOE) prepared this programmatic environmental impact statement to assess the potential impacts of alternative management strategies for depleted uranium hexafluoride currently stored at three DOE sites: Paducah site near Paducah, Kentucky; Portsmouth site near Portsmouth, Ohio; and K-25 site on the Oak Ridge Reservation in Oak Ridge, Tennessee.

30

Characterization of Thermal Properties of Depleted Uranium Metal Microspheres  

E-Print Network [OSTI]

that combines these previous two methods to characterize the diffusivity of a packed bed of microspheres of depleted uranium (DU) metal, which have a nominal diameter of 250 micrometers. The new apparatus is designated as the Crucible Heater Test Assembly (CHTA...

Humrickhouse, Carissa Joy

2012-07-16T23:59:59.000Z

31

Sampling Plan for Assaying Plates Containing Depleted or Normal Uranium  

SciTech Connect (OSTI)

This paper describes the rationale behind the proposed method for selecting a 'representative' sample of uranium metal plates, portions of which will be destructively assayed at the Y-12 Security Complex. The total inventory of plates is segregated into two populations, one for Material Type 10 (depleted uranium (DU)) and one for Material Type 81 (normal [or natural] uranium (NU)). The plates within each population are further stratified by common dimensions. A spreadsheet gives the collective mass of uranium element (and isotope for DU) and the piece count of all plates within each stratum. These data are summarized in Table 1. All plates are 100% uranium metal, and all but approximately 60% of the NU plates have Kel-F{reg_sign} coating. The book inventory gives an overall U-235 isotopic percentage of 0.22% for the DU plates, ranging from 0.19% to 0.22%. The U-235 ratio of the NU plates is assumed to be 0.71%. As shown in Table 1, the vast majority of the plates are comprised of depleted uranium, so most of the plates will be sampled from the DU population.

Ivan R. Thomas

2011-11-01T23:59:59.000Z

32

Ultrafiltration evaluation with depleted uranium oxide  

SciTech Connect (OSTI)

Scientists at the Los Alamos National Laboratory Plutonium Facility are using electrodissolution in neutral to alkaline solutions to decontaminate oralloy parts that have surface plutonium contamination. Ultrafiltration of the electrolyte stream removes precipitate so that the electrolyte stream to the decontamination fixture is precipitate free. This report describes small-scale laboratory ultrafiltration experiments that the authors performed to determine conditions necessary for full-scale operation of an ultrafiltration module. Performance was similar to what they observed in the ferric hydroxide system. At 12 psi transmembrane pressure, a shear rate of 12,000 sec{sup {minus}1} was sufficient to sustain membrane permeability. Ultrafiltration of uranium(VI) oxide appears to occur as easily as ultrafiltration of ferric hydroxide. Considering the success reported in this study, the authors plan to add ultrafiltration to the next decontamination system for oralloy parts.

Weisbrod, K.R.; Schake, A.R.; Morgan, A.N.; Purdy, G.M.; Martinez, H.E.; Nelson, T.O.

1998-03-01T23:59:59.000Z

33

Microstructure of depleted uranium under uniaxial strain conditions  

SciTech Connect (OSTI)

Uranium samples of two different purities were used for spall strength measurements. Samples of depleted uranium were taken from very high purity material (38 ppM carbon) and from material containing 280 ppM C. Experimental conditions were chosen to effectively arrest the microstructural damage at two places in the development to full spall separation. Samples were soft recovered and characterized with respect to the microstructure and the form of damage. This allowed determination of the dependence of spall mechanisms on stress level, stress state, and sample purity. This information is used in developing a model to predict the mode of fracture.

Zurek, A.K.; Embury, J.D.; Kelly, A.; Thissell, W.R.; Gustavsen, R.L.; Vorthman, J.E.; Hixson, R.H.

1997-09-01T23:59:59.000Z

34

Uranio impoverito: perch'e? (Depleted uranium: why?)  

E-Print Network [OSTI]

In this paper we develop a simple model of the penetration process of a long rod through an uniform target. Applying the momentum and energy conservation laws, we derive an analytical relation which shows how the penetration depth depends upon the density of the rod, given a fixed kinetic energy. This work was sparked off by the necessity of the author of understanding the reasons of the effectiveness of high density penetrators (e.g. depleted uranium penetrators) as anti-tank weapons.

D'Abramo, G

2003-01-01T23:59:59.000Z

35

Depleted uranium storage and disposal trade study: Summary report  

SciTech Connect (OSTI)

The objectives of this study were to: identify the most desirable forms for conversion of depleted uranium hexafluoride (DUF6) for extended storage, identify the most desirable forms for conversion of DUF6 for disposal, evaluate the comparative costs for extended storage or disposal of the various forms, review benefits of the proposed plasma conversion process, estimate simplified life-cycle costs (LCCs) for five scenarios that entail either disposal or beneficial reuse, and determine whether an overall optimal form for conversion of DUF6 can be selected given current uncertainty about the endpoints (specific disposal site/technology or reuse options).

Hightower, J.R.; Trabalka, J.R.

2000-02-01T23:59:59.000Z

36

EIS-0329: Proposed Construction, Operation, Decontamination/Decommissioning of Depleted Uranium Hexafluoride Conversion Facilities  

Broader source: Energy.gov [DOE]

This EIS analyzes DOE's proposal to construct, operate, maintain, and decontaminate and decommission two depleted uranium hexafluoride (DUF 6) conversion facilities, at Portsmouth, Ohio, and Paducah, Kentucky.

37

Including environmental concerns in management strategies for depleted uranium hexafluoride  

SciTech Connect (OSTI)

One of the major programs within the Office of Nuclear Energy, Science, and Technology of the US Department of Energy (DOE) is the depleted uranium hexafluoride (DUF{sub 6}) management program. The program is intended to find a long-term management strategy for the DUF{sub 6} that is currently stored in approximately 46,400 cylinders at Paducah, KY; Portsmouth, OH; and Oak Ridge, TN, USA. The program has four major components: technology assessment, engineering analysis, cost analysis, and the environmental impact statement (EIS). From the beginning of the program, the DOE has incorporated the environmental considerations into the process of strategy selection. Currently, the DOE has no preferred alternative. The results of the environmental impacts assessment from the EIS, as well as the results from the other components of the program, will be factored into the strategy selection process. In addition to the DOE`s current management plan, other alternatives continued storage, reuse, or disposal of depleted uranium, will be considered in the EIS. The EIS is expected to be completed and issued in its final form in the fall of 1997.

Goldberg, M. [Argonne National Laboratory, Washington, DC (United States); Avci, H.I. [Argonne National Lab., IL (United States); Bradley, C.E. [USDOE, Washington, DC (United States)

1995-12-31T23:59:59.000Z

38

Depleted-Uranium Weapons the Whys and Wherefores  

E-Print Network [OSTI]

The only military application in which present-day depleted-uranium (DU) alloys out-perform tungsten alloys is long-rod penetration into a main battle-tank's armor. However, this advantage is only on the order of 10% and disappearing when the comparison is made in terms of actual lethality of complete anti-tank systems instead of laboratory-type steel penetration capability. Therefore, new micro- and nano-engineered tungsten alloys may soon out-perform existing DU alloys, enabling the production of tungsten munition which will be better than uranium munition, and whose overall life-cycle cost will be less due to the absence of the problems related to the radioactivity of uranium. The reasons why DU weapons have been introduced and used are analysed from the perspective that their radioactivity must have played an important role in the decision making process. It is found that DU weapons belong to the diffuse category of low-radiological-impact nuclear weapons to which emerging types of low-yield, i.e., fourth...

Gsponer, A

2003-01-01T23:59:59.000Z

39

Military use of depleted uranium assessment of prolonged population exposure  

E-Print Network [OSTI]

This work is an exposure assessment for a population living in an area contaminated by use of depleted uranium (DU) weapons. RESRAD 5.91 code is used to evaluate the average effective dose delivered from 1, 10, 20 cm depths of contaminated soil, in a residential farmer scenario. Critical pathway and group are identified in soil inhalation or ingestion and children playing with the soil, respectively. From available information on DU released on targeted sites, both critical and average exposure can leave to toxicological hazards; annual dose limit for population can be exceeded on short-term period (years) for soil inhalation. As a consequence, in targeted sites cleaning up must be planned on the basis of measured concentration, when available, while special cautions have to be adopted altogether to reduce unaware exposures, taking into account the amount of the avertable dose.

Giannardi, C

2001-01-01T23:59:59.000Z

40

FEASIBILITY STUDY OF DUPOLY TO RECYCLE DEPLETED URANIUM.  

SciTech Connect (OSTI)

DUPoly, depleted uranium (DU) powder microencapsulated in a low-density polyethylene binder, has been demonstrated as an innovative and efficient recycle product, a very durable high density material with significant commercial appeal. DUPoly was successfully prepared using uranium tetrafluoride (UF{sub 4}) ''green salt'' obtained from Fluor Daniel-Fernald, a U.S. Department of Energy reprocessing facility near Cincinnati, Ohio. Samples containing up to 90 wt% UF{sub 4} were produced using a single screw plastics extruder, with sample densities of up to 3.97 {+-} 0.08 g/cm{sup 3} measured. Compressive strength of as-prepared samples (50-90 wt% UF4 ) ranged from 1682 {+-} 116 psi (11.6 {+-} 0.8 MPa) to 3145 {+-} 57 psi (21.7 {+-} 0.4 MPa). Water immersion testing for a period of 90 days produced no visible degradation of the samples. Leach rates were low, ranging from 0.02 % (2.74 x 10{sup {minus}6} gm/gm/d) for 50 wt% UF{sub 4} samples to 0.72 % (7.98 x 10{sup {minus}5} gm/gm/d) for 90 wt% samples. Sample strength was not compromised by water immersion. DUPoly samples containing uranium trioxide (UO{sub 3}), a DU reprocessing byproduct material stockpiled at the Savannah River Site, were gamma irradiated to 1 x 10{sup 9} rad with no visible deterioration. Compressive strength increased significantly, however: up to 200% for samples with 90 wt% UO{sub 3}. Correspondingly, percent deformation (strain) at failure was decreased for all samples. Gamma attenuation data on UO{sub 3} DUPoly samples yielded mass attenuation coefficients greater than those for lead. Neutron removal coefficients were calculated and shown to correlate well with wt% of DU. Unlike gamma attenuation, both hydrogenous and nonhydrogenous materials interact to attenuate neutrons.

ADAMS,J.W.; LAGERAAEN,P.R.; KALB,P.D.; RUTENKROGER,S.P.

1998-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Attainable Burnup in a LIFE Engine Loaded with Depleted Uranium  

SciTech Connect (OSTI)

The Laser Inertial Fusion-based Energy (LIFE) system uses a laser-based fusion source for electricity production. The (D,T) reaction, beside a pure fusion system, allows the option to drive a sub-critical fission blanket in order to increase the total energy gain. In a typical fusion-fission LIFE engine the fission blanket is a spherical shell around the fusion source, preceded by a beryllium shell for neutron multiplications by means of (n,2n) reactions. The fuel is in the form of TRISO particles dispersed in carbon pebbles, cooled by flibe. The optimal design features 80 cm thick blanket, 16 cm multiplier, and 20% TRISO packing factor. A blanket loaded with depleted uranium and depleted in a single batch with continuous mixing can achieve burnup as high as {approx}85% FIMA while generating 2,000 MW of total thermal power and producing enough tritium to be used for fusion. A multi-segment blanket with a central promotion shuffling scheme enhances burnup to {approx}90% FIMA, whereas a blanket that is operated with continuous refueling achieves only 82% FIMA under the same constraints of thermal power and tritium self-sufficiency. Both, multi-segment and continuous refueling eliminate the need for a fissile breeding phase.

Fratoni, M; Kramer, K J; Latkowski, J F

2009-11-30T23:59:59.000Z

42

Proposal concerning the participation of CERN in the procurement of depleted-uranium sheets for the UA1 calorimeter upgrading  

E-Print Network [OSTI]

Proposal concerning the participation of CERN in the procurement of depleted-uranium sheets for the UA1 calorimeter upgrading

1985-01-01T23:59:59.000Z

43

Delayed neutron measurements for Th-232, Np-237, Pu-239, Pu-241 and depleted uranium  

E-Print Network [OSTI]

The neutron emission rates from five very pure actinide samples (Th-232, Np-237, Pu-239, Pu-241 and depleted uranium) were measured following equilibrium irradiation in fast and thermal neutron fluxes. The relative abundances (alphas) for the first...

Stone, Joseph C.

2001-01-01T23:59:59.000Z

44

Composition of the U.S. DOE Depleted Uranium Inventory  

E-Print Network [OSTI]

about 2.75 wt% U-235. For further enrichment, the material was shipped to the Oak Ridge and Portsmouth plants. In addition to natural uranium, also uranium recycled from spent fuel was fed into the Paducah enrichment cascade (Table 2 and Fig. 2). The recycled uranium introduced various isotopes not found in natural uranium into the cascade: fission products, such as Technetium-99; transuranics, such as Neptunium-237 and Plutonium-239; and the artificial uranium isotope of Uranium-236. The spent fuel, from which uranium was recycled, originated from the Hanford and Savannah River military plutonium production reactors. This uranium was recycled, although its assay of U-235 was somewhat lower than in natural uranium (Table 2). This obviously must be seen in the context of the Cold War era, when uranium was a scarce resource. Due to the low burn-up of the military reactors, concentrations of artificial U-236 are comparatively low in this recycled uranium. The recycled uranium represents

Concentration Of Less

45

DUSCOBS - a depleted-uranium silicate backfill for transport, storage, and disposal of spent nuclear fuel  

SciTech Connect (OSTI)

A Depleted Uranium Silicate COntainer Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside storage, transport, and repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill all void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (1) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (2) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. In addition, the DUSCOBS improves the integrity of the package by acting as a packing material and ensures criticality control for the package during SNF storage and transport. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments.

Forsberg, C.W.; Pope, R.B.; Ashline, R.C.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

1995-11-30T23:59:59.000Z

46

Summary of the engineering analysis report for the long-term management of depleted uranium hexafluoride  

SciTech Connect (OSTI)

The Department of Energy (DOE) is reviewing ideas for the long-term management and use of its depleted uranium hexafluoride. DOE owns about 560,000 metric tons (over a billion pounds) of depleted uranium hexafluoride. This material is contained in steel cylinders located in storage yards near Paducah, Kentucky; Portsmouth, Ohio; and at the East Tennessee Technology Park (formerly the K-25 Site) in Oak Ridge, Tennessee. On November 10, 1994, DOE announced its new Depleted Uranium Hexafluoride Management Program by issuing a Request for Recommendations and an Advance Notice of Intent in the Federal Register (59 FR 56324 and 56325). The first part of this program consists of engineering, costs and environmental impact studies. Part one will conclude with the selection of a long-term management plan or strategy. Part two will carry out the selected strategy.

Dubrin, J.W., Rahm-Crites, L.

1997-09-01T23:59:59.000Z

47

A comparison of two lung clearance models based on the dissolution rates of oxidized depleted uranium  

E-Print Network [OSTI]

by Cuddihy. Predictions fr'om bai. h models based on the dissolution rates of the amount of oxidized depleted uranium that wau'ld be cleared to blood irom the pu lraana ry region i'o'i)owing an i nba !at i cn exposure were compared . It was f:urd ti... to oxidized depleted uranium (DU) aerosol. The ob, ject. ive of th. is i:hesis was three fold: (1) to determine the dissolution rates for two respirable DU samples, (2) to determine the specific pulmonary clearance characteristics of oxidized DU, (3) Co...

Crist, Kevin Craig

1983-01-01T23:59:59.000Z

48

Depleted Uranium Hexafluoride Management Program. The technology assessment report for the long-term management of depleted uranium hexafluoride. Volume 1  

SciTech Connect (OSTI)

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

Zoller, J.N.; Rosen, R.S.; Holliday, M.A. [and others] [and others

1995-06-30T23:59:59.000Z

49

Depleted Uranium Hexafluoride Management Program. The technology assessment report for the long-term management of depleted uranium hexafluoride. Volume 2  

SciTech Connect (OSTI)

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

Zoller, J.N.; Rosen, R.S.; Holliday, M.A. [and others] [and others

1995-06-30T23:59:59.000Z

50

D0 Decomissioning : Storage of Depleted Uranium Modules Inside D0 Calorimeters after the Termination of D0 Experiment  

SciTech Connect (OSTI)

Dzero liquid Argon calorimeters contain hadronic modules made of depleted uranium plates. After the termination of DO detector's operation, liquid Argon will be transferred back to Argon storage Dewar, and all three calorimeters will be warmed up. At this point, there is no intention to disassemble the calorimeters. The depleted uranium modules will stay inside the cryostats. Depleted uranium is a by-product of the uranium enrichment process. It is slightly radioactive, emits alpha, beta and gamma radiation. External radiation hazards are minimal. Alpha radiation has no external exposure hazards, as dead layers of skin stop it; beta radiation might have effects only when there is a direct contact with skin; and gamma rays are negligible - levels are extremely low. Depleted uranium is a pyrophoric material. Small particles (such as shavings, powder etc.) may ignite with presence of Oxygen (air). Also, in presence of air and moisture it can oxidize. Depleted uranium can absorb moisture and keep oxidizing later, even after air and moisture are excluded. Uranium oxide can powder and flake off. This powder is also pyrographic. Uranium oxide may create health problems if inhaled. Since uranium oxide is water soluble, it may enter the bloodstream and cause toxic effects.

Sarychev, Michael; /Fermilab

2011-09-21T23:59:59.000Z

51

Composition for radiation shielding  

DOE Patents [OSTI]

A composition for use as a radiation shield. The shield has a depleted urum core for absorbing gamma rays and a bismuth coating for preventing chemical corrosion and absorbing gamma rays. Alternatively, a sheet of gadolinium may be positioned between the uranium core and the bismuth coating for absorbing neutrons. The composition is preferably in the form of a container for storing materials that emit radiation such as gamma rays and neutrons. The container is preferably formed by casting bismuth around a pre-formed uranium container having a gadolinium sheeting, and allowing the bismuth to cool. The resulting container is a structurally sound, corrosion-resistant, radiation-absorbing container.

Kronberg, James W. (Aiken, SC)

1994-01-01T23:59:59.000Z

52

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

Broader source: Energy.gov [DOE]

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 Paducah site; transportation of all cylinders (DUF6, enriched, and empty) currently stored at the East Tennessee Technology Park (ETTP) near Oak Ridge, Tennessee, to Portsmouth; construction of a new cylinder storage yard at Portsmouth (if required) for ETTP cylinders; transportation of depleted uranium conversion products and waste materials to a disposal facility; transportation and sale of the hydrogen fluoride (HF) produced as a conversion coproduct; and neutralization of HF to calcium fluoride and its sale or disposal in the event that the HF product is not sold.

53

Process for producing an aggregate suitable for inclusion into a radiation shielding product  

DOE Patents [OSTI]

The present invention is directed to methods for converting depleted uranium hexafluoride to a stable depleted uranium silicide in a one-step reaction. Uranium silicide provides a stable aggregate material that can be added to concrete to increase the density of the concrete and, consequently, shield gamma radiation. As used herein, the term "uranium silicide" is defined as a compound generically having the formula U.sub.x Si.sub.y, wherein the x represents the molecules of uranium and the y represent the molecules of silicon. In accordance with the present invention, uranium hexafluoride is converted to a uranium silicide by contacting the uranium hexafluoride with a silicon-containing material at a temperature in a range between about 1450.degree. C. and about 1750.degree. C. The stable depleted uranium silicide is included as an aggregate in a radiation shielding product, such as a concrete product.

Lessing, Paul A. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID)

2000-01-01T23:59:59.000Z

54

Assessing the risk from the depleted uranium weapons used in Operation Allied Force  

E-Print Network [OSTI]

The conflict in Yugoslavia has been a source of great concern for the neighboring countries, about the radiological and toxic hazard posed by the alleged presence of depleted uranium in NATO weapons. In the present study a worst-case scenario is assumed mainly to assess the risk for Greece and other neighboring countries of Yugoslavia at similar distances . The risk of the weapons currently in use is proved to be negligible at distances greater than 100 Km. For shorter distances classified data of weapons composition are needed to obtain a reliable assessment.

Liolios, T E

1999-01-01T23:59:59.000Z

55

Proposal for the award of a contract for the supply of 5 mm depleted-uranium plates for the UA1 calorimeter upgrading  

E-Print Network [OSTI]

Proposal for the award of a contract for the supply of 5 mm depleted-uranium plates for the UA1 calorimeter upgrading

1986-01-01T23:59:59.000Z

56

Proposal for the award of a contract for the supply of 5 mm depleted-uranium plates for the UA1 experiment  

E-Print Network [OSTI]

Proposal for the award of a contract for the supply of 5 mm depleted-uranium plates for the UA1 experiment

1986-01-01T23:59:59.000Z

57

Depleted uranium hexafluoride (DUF{sub 6}) management system--a decision tool  

SciTech Connect (OSTI)

The Depleted Uranium Hexafluoride (DUF{sub 6}) Management System (DMS) is being developed as a decision tool to provide cost and risk data for evaluation of short-and long-term management strategies for depleted uranium. It can be used to assist decision makers on a programmatic or site-specific level. Currently, the DMS allows evaluation of near-term cylinder management strategies such as storage yard improvements, cylinder restocking, and reconditioning. The DMS has been designed to provide the user with maximum flexibility for modifying data and impact factors (e.g., unit costs and risk factors). Sensitivity analysis can be performed on all key parameters such as cylinder corrosion rate, inspection frequency, and impact factors. Analysis may be conducted on a system-wide, site, or yard basis. The costs and risks from different scenarios may be compared in graphic or tabular format. Ongoing development of the DMS will allow similar evaluation of long-term management strategies such as conversion to other chemical forms. The DMS is a Microsoft Windows 3.1 based, stand-alone computer application. It can be operated on a 486 or faster computer with VGA, 4 MB of RAM, and 10 MB of disk space.

Gasper, J.R.; Sutter, R.J.; Avci, H.I. [and others

1995-12-31T23:59:59.000Z

58

Summary of the cost analysis report for the long-term management of depleted uranium hexafluoride  

SciTech Connect (OSTI)

This report is a summary of the Cost Analysis Report which provides comparative cost data for the management strategy alternatives. The PEIS and the Cost Analysis Report will help DOE select a management strategy. The Record of Decision, expected in 1998, will complete the first part of the Depleted Uranium Hexafluoride Management Program. The second part of the Program will look at specific sites and technologies for carrying out the selected strategy. The Cost Analysis Report estimates the primary capital and operating costs for the different alternatives. It reflects the costs of technology development construction of facilities, operation, and decontamination and decommissioning. It also includes potential revenues from the sale of by-products such as anhydrous hydrogen fluoride (ABF). These estimates are based on early designs. They are intended to help in comparing alternatives, rather than to indicate absolute costs for project budgets or bidding purposes. More detailed estimates and specific funding sources will be considered in part two of the Depleted Uranium Hexafluoride Management Program.

Dubrin, J.W.; Rahm-Crites, L.

1997-09-01T23:59:59.000Z

59

Composition for radiation shielding  

DOE Patents [OSTI]

A composition for use as a radiation shield is disclosed. The shield has a depleted uranium core for absorbing gamma rays and a bismuth coating for preventing chemical corrosion and absorbing gamma rays. Alternatively, a sheet of gadolinium may be positioned between the uranium core and the bismuth coating for absorbing neutrons. The composition is preferably in the form of a container for storing materials that emit radiation such as gamma rays and neutrons. The container is preferably formed by casting bismuth around a pre-formed uranium container having a gadolinium sheeting, and allowing the bismuth to cool. The resulting container is a structurally sound, corrosion-resistant, radiation-absorbing container. 2 figs.

Kronberg, J.W.

1994-08-02T23:59:59.000Z

60

Using Hydro-Cutting to Aid in Remediation of a Firing Range Contaminated with Depleted Uranium  

SciTech Connect (OSTI)

This paper describes the challenges encountered in decommissioning a firing range that had been used to test fire depleted uranium rounds in the late 1950's and early 1960's. The paper details the operational challenges and innovative solutions involved in remediating and decommissioning a firing range bullet catcher once unexploded ordnance was discovered. It also discusses how the Army dealt with an intertwining web of regulatory and permit issues that arose in treating and disposing of multiple waste streams. The paper will show how the use of a Resource Conservation and Recovery Act (RCRA) Temporary Authorization allowed the Army to deal with the treatment of a variety of waste streams and how hydro-cutting process was used to demilitarize the potentially unexploded rounds.

Styvaert, Michael S.; Conley, Richard D.; Watters, David J.

2003-02-24T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Remediation application strategies for depleted uranium contaminated soils at the US Army Yuma Proving Ground  

SciTech Connect (OSTI)

The US Army Yuma Proving Ground (YPG), located in the southwest portion of Arizona conducts firing of projectiles into the Gunpoint (GP-20) firing range. The penetrators are composed of titanium and DU. The purpose of this project was to determine feasible cleanup technologies and disposal alternatives for the cleanup of the depleted uranium (DU) contaminated soils at YPG. The project was split up into several tasks that include (a) collecting and analyzing samples representative of the GP-20 soils, (b) evaluating the data results, (c) conducting a literature search of existing proven technologies for soil remediation, and (0) making final recommendations for implementation of this technology to the site. As a result of this study, several alternatives for the separation, treatment, and disposal procedures are identified that would result in meeting the cleanup levels defined by the Nuclear Regulatory Commission for unrestricted use of soils and would result in a significant cost savings over the life of the firing range.

Vandel, D.S.; Medina, S.M.; Weidner, J.R.

1994-03-01T23:59:59.000Z

62

ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.  

SciTech Connect (OSTI)

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specificationsa and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 11 began in early January 1958, and the Assembly 11 program ended in late January 1958. The core consisted of highly enriched uranium (HEU) plates and depleted uranium plates loaded into stainless steel drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, six columns of 0.125 in.-wide (3.175 mm) depleted uranium plates and one column of 1.0 in.-wide (25.4 mm) depleted uranium plates. The length of each column was 10 in. (254.0 mm) in each half of the core. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the depleted uranium radial blanket was approximately 14 in. (355.6 mm), and the length of the radial blanket in each half of the matrix was 22 in. (558.8 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/11, the reference critical configuration was loading 10 which was critical on January 21, 1958. Subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/11 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/11 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertain

Lell, R. M.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; National Security; Inst. of Physics and Power Engineering

2010-09-30T23:59:59.000Z

63

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

SciTech Connect (OSTI)

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

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

2003-02-27T23:59:59.000Z

64

Parametric Evaluation of Active Neutron Interrogation for the Detection of Shielded Highly-Enriched Uranium in the Field  

SciTech Connect (OSTI)

Parametric studies using numerical simulations are being performed to assess the performance capabilities and limits of active neutron interrogation for detecting shielded highly enriched uranium (HEU). Varying the shield material, HEU mass, HEU depth inside the shield, and interrogating neutron source energy, the simulations account for both neutron and photon emission signatures from the HEU with resolution in both energy and time. The results are processed to represent different irradiation timing schemes and several different classes of radiation detectors, and evaluated using a statistical approach considering signal intensity over background. This paper describes the details of the modeling campaign and some preliminary results, weighing the strengths of alternative measurement approaches for the different irradiation scenarios.

D. L. Chcihester; E. H. Seabury; S. J. Thompson; R. R. C. Clement

2011-10-01T23:59:59.000Z

65

Modeling exposure to depleted uranium in support of decommissioning at Jefferson Proving Ground, Indiana  

SciTech Connect (OSTI)

Jefferson Proving Ground was used by the US Army Test and Evaluation Command for testing of depleted uranium munitions and closed in 1995 under the Base Realignment and Closure Act. As part of the closure of JPG, assessments of potential adverse health effects to humans and the ecosystem were conducted. This paper integrates recent information obtained from site characterization surveys at JPG with environmental monitoring data collected from 1983 through 1994 during DU testing. Three exposure scenarios were evaluated for potential adverse effects to human health: an occasional use scenario and two farming scenarios. Human exposure was minimal from occasional use, but significant risk were predicted from the farming scenarios when contaminated groundwater was used by site occupants. The human health risk assessments do not consider the significant risk posed by accidents with unexploded ordnance. Exposures of white-tailed deer to DU were also estimated in this study, and exposure rates result in no significant increase in either toxicological or radiological risks. The results of this study indicate that remediation of the DU impact area would not substantially reduce already low risks to humans and the ecosystem, and that managed access to JPG is a reasonable model for future land use options.

Ebinger, M.H. [Los Alamos National Lab., NM (United States); Oxenburg, T.P. [Army Test and Evaluation Command, Aberdeen Proving Ground, MD (United States)

1997-02-01T23:59:59.000Z

66

Proceedings of a workshop on uses of depleted uranium in storage, transportation and repository facilities  

SciTech Connect (OSTI)

A workshop on the potential uses of depleted uranium (DU) in the repository was organized to coordinate the planning of future activities. The attendees, the original workshop objective and the agenda are provided in Appendices A, B and C. After some opening remarks and discussions, the objectives of the workshop were revised to: (1) exchange information and views on the status of the Department of Energy (DOE) activities related to repository design and planning; (2) exchange information on DU management and planning; (3) identify potential uses of DU in the storage, transportation, and disposal of high-level waste and spent fuel; and (4) define the future activities that would be needed if potential uses were to be further evaluated and developed. This summary of the workshop is intended to be an integrated resource for planning of any future work related to DU use in the repository. The synopsis of the first day`s presentations is provided in Appendix D. Copies of slides from each presenter are presented in Appendix E.

NONE

1997-12-31T23:59:59.000Z

67

Advancing Performance Assessment for Disposal of Depleted Uranium at Clive Utah - 12493  

SciTech Connect (OSTI)

A Performance Assessment (PA) for disposal of depleted uranium (DU) waste has recently been completed for a potential disposal facility at Clive in northwestern Utah. For the purposes of this PA, 'DU waste' includes uranium oxides of all naturally-occurring isotopes, though depleted in U-235, varying quantities of other radionuclides introduced to the uranium enrichment process in the form of used nuclear reactor fuel (reactor returns), and decay products of all of these radionuclides. The PA will be used by the State of Utah to inform an approval decision for disposal of DU waste at the facility, and will be available to federal regulators as they revisit rulemaking for the disposal of DU. The specific performance objectives of the Clive DU PA relate to annual individual radiation dose within a 10,000-year performance period, groundwater concentrations of specific radionuclides within a 500-year compliance period, and site stability in the longer term. Fate and transport processes that underlie the PA model include radioactive decay and ingrowth, diffusion in gaseous and water phases, water advection in unsaturated and saturated zones, transport caused by plant and animal activity, cover naturalization, natural and anthropogenic erosion, and air dispersion. Fate and transport models were used to support the dose assessment and the evaluation of groundwater concentrations. Exposure assessment was based on site-specific scenarios, since the traditional human exposure scenarios suggested by DOE and NRC guidance are unrealistic for this site. Because the U-238 in DU waste reaches peak radioactivity (secular equilibrium) after 2 million years (My) following its separation, the PA must also evaluate the impact of climate change cycles, including the return of pluvial lakes such as Lake Bonneville. The first draft of the PA has been submitted to the State of Utah for review. The results of this preliminary analysis indicate that doses are very low for the site-specific receptors for the 10,000-year compliance period. This is primarily because DU waste is not highly radioactive within this time frame, the DU waste is assumed to be buried beneath zones exposed by erosion, groundwater concentrations of DU waste constituents do not exceed groundwater protection limits with in the 500-year compliance period, and the first deep lake occurrence will disperse DU waste across a large area, and will ultimately be covered by lake-derived sediment. A probabilistic PA model was constructed that considered DU waste and decay product doses to site-specific receptors for a 10,000-yr performance period, as well as deep-time effects. The quantitative results are summarized in Table VII. Doses (as TEDE) are always less than 5 mSv in a year, and doses to the offsite receptors are always much less than 0.25 mSv in a year. Groundwater concentrations of Tc-99 are always less than its GWPL except when the Tc-99 contaminated waste is disposed below grade. Even in this case, the median groundwater concentration is only 4.18 Bq/L (113 pCi/L), which is more than one order of magnitude less than the GWPL for Tc-99. The results overall suggest that there are disposal configurations that can be used to dispose of the proposed quantities of DU waste that are adequately protective of human health. (authors)

Black, Paul; Tauxe, John; Perona, Ralph; Lee, Robert; Catlett, Kate; Balshi, Mike; Fitzgerald, Mark; McDermott, Greg [Neptune and Company, Inc., Los Alamos, New Mexico 87544 (United States); Shrum, Dan; McCandless, Sean; Sobocinski, Robert; Rogers, Vern [EnergySolutions, LLC, Salt Lake City, Utah 84101 (United States)

2012-07-01T23:59:59.000Z

68

Determination of Depleted Uranium in Environmental Bio-monitor Samples and Soil from Target sites in Western Balkan Region  

SciTech Connect (OSTI)

Lichen and Moss are widely used to assess the atmospheric pollution by heavy metals and radionuclides. In this paper, we report results of uranium and its isotope ratios using mass spectrometric measurements (followed by chemical separation procedure) for mosses, lichens and soil samples from a depleted uranium (DU) target site in western Balkan region. Samples were collected in 2003 from Han Pijesak (Republika Srpska in Bosnia and Hercegovina). Inductively coupled plasma mass spectrometry (ICP-MS) measurements show the presence of high concentration of uranium in some samples. Concentration of uranium in moss samples ranged from 5.2-755.43 Bq/Kg. We have determined {sup 235}U/{sup 238}U isotope ratio using thermal ionization mass spectrometry (TIMS) from the samples with high uranium content and the ratios are in the range of 0.002097-0.002380. TIMS measurement confirms presence of DU in some samples. However, we have not noticed any traces of DU in samples containing lesser amount of uranium or from any samples from the living environment of same area.

Sahoo, Sarata K.; Enomoto, Hiroko; Tokonami, Shinji; Ishikawa, Tetsuo [National Institute of Radiological Sciences, 4-9-1 Anagawa, Inage-ku, Chiba 263-8555 (Japan); Ujic, Predrag; Celikovic, Igor; Zunic, Zora S. [Institute of Nuclear Sciences, Vinca, Mike Petrovica Alasa 12-14, 11000 Belgrade (Serbia)

2008-08-07T23:59:59.000Z

69

Determination of Young's modulus and mechanical damping as a function of temperature for depleted uranium-0.75 wt% titanium using the PUCOT  

E-Print Network [OSTI]

DETERMINATION OF YOUNG'S MODULUS AND MECHANICAL DAMPING AS A FUNCTION OF TEMPERATURE FOR DEPLETED URANIUM-0. 75 WT% TITANIUM USING THE PUCOT A Thesis bY KEITH HOWARD KEENE Submitted to the Graduate College of Texas A&M University in partial... fulfillment of requirements for the degree of MASTER OF SCIENCE May 1986 Major Subject: Mechanical Engineerinq DETERMINATION OF YOUNG'S MODULUS AND MECHANICAL DAMPING AS A FUNCTION OF TEMPERATURE FOR DEPLETED URANIUM-0. 75 WT% TITANIUM USING THE PUCOT A...

Keene, Keith Howard

1986-01-01T23:59:59.000Z

70

Bacterial Community Succession During in situ Uranium Bioremediation: Spatial Similarities Along Controlled Flow Paths  

E-Print Network [OSTI]

problem, and the use of depleted uranium and other heavyenvironmental hazard. Depleted uranium is weakly radioactive

Hwang, Chiachi

2009-01-01T23:59:59.000Z

71

Evaluation of depleted uranium in the environment at Aberdeen Proving Grounds, Maryland and Yuma Proving Grounds, Arizona. Final report  

SciTech Connect (OSTI)

This report represents an evaluation of depleted uranium (DU) introduced into the environment at the Aberdeen Proving Grounds (APG), Maryland and Yuma Proving Grounds (YPG) Arizona. This was a cooperative project between the Environmental Sciences and Statistical Analyses Groups at LANL and with the Department of Fishery and Wildlife Biology at Colorado State University. The project represents a unique approach to assessing the environmental impact of DU in two dissimilar ecosystems. Ecological exposure models were created for each ecosystem and sensitivity/uncertainty analyses were conducted to identify exposure pathways which were most influential in the fate and transport of DU in the environment. Research included field sampling, field exposure experiment, and laboratory experiments. The first section addresses DU at the APG site. Chapter topics include bioenergetics-based food web model; field exposure experiments; bioconcentration by phytoplankton and the toxicity of U to zooplankton; physical processes governing the desorption of uranium from sediment to water; transfer of uranium from sediment to benthic invertebrates; spead of adsorpion by benthic invertebrates; uptake of uranium by fish. The final section of the report addresses DU at the YPG site. Chapters include the following information: Du transport processes and pathway model; field studies of performance of exposure model; uptake and elimination rates for kangaroo rates; chemical toxicity in kangaroo rat kidneys.

Kennedy, P.L.; Clements, W.H.; Myers, O.B.; Bestgen, H.T.; Jenkins, D.G. [Colorado State Univ., Fort Collins, CO (United States). Dept. of Fishery and Wildlife Biology

1995-01-01T23:59:59.000Z

72

Long-term criticality control in radioactive waste disposal facilities using depleted uranium  

SciTech Connect (OSTI)

Plant photosynthesis has created a unique planetary-wide geochemistry - an oxidizing atmosphere with oxidizing surface waters on a planetary body with chemically reducing conditions near or at some distance below the surface. Uranium is four orders of magnitude more soluble under chemically oxidizing conditions than it is under chemically reducing conditions. Thus, uranium tends to leach from surface rock and disposal sites, move with groundwater, and concentrate where chemically reducing conditions appear. Earth`s geochemistry concentrates uranium and can separate uranium from all other elements except oxygen, hydrogen (in water), and silicon (silicates, etc). Fissile isotopes include {sup 235}U, {sup 233}U, and many higher actinides that eventually decay to one of these two uranium isotopes. The potential for nuclear criticality exists if the precipitated uranium from disposal sites has a significant fissile enrichment, mass, and volume. The earth`s geochemistry suggests that isotopic dilution of fissile materials in waste with {sup 238}U is a preferred strategy to prevent long-term nuclear criticality in and beyond the boundaries of waste disposal facilities because the {sup 238}U does not separate from the fissile uranium isotopes. Geological, laboratory, and theoretical data indicate that the potential for nuclear criticality can be minimized by diluting fissile materials with-{sup 238}U to 1 wt % {sup 235}U equivalent.

Forsberg, C.W.

1997-02-19T23:59:59.000Z

73

ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.  

SciTech Connect (OSTI)

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3 Assembly 6F (ZPR-3/6F), the final phase of the Assembly 6 program, simulated a spherical core with a thick depleted uranium reflector. ZPR-3/6F was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 47 at.%. Approximately 81.4% of the total fissions in this assembly occur above 100 keV, approximately 18.6% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 7 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3/6F began in late December 1956, and the experimental measurements were performed in January 1957. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates, perforated aluminum plates and stainless steel plates loaded into aluminum drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of three columns of 0.125 in.-wide (3.175 mm) HEU plates, three columns of 0.125 in.-wide depleted uranium plates, nine columns of 0.125 in.-wide perforated aluminum plates and one column of stainless steel plates. The maximum length of each column of core material in a drawer was 9 in. (228.6 mm). Because of the goal to produce an approximately spherical core, core fuel and diluent column lengths generally varied between adjacent drawers and frequently within an individual drawer. The axial reflector consisted of depleted uranium plates and blocks loaded in the available space in the front (core) drawers, with the remainder loaded into back drawers behind the front drawers. The radial reflector consisted of blocks of depleted uranium loaded directly into the matrix tubes. The assembly geometry approximated a reflected sphere as closely as the square matrix tubes, the drawers and the shapes of fuel and diluent plates allowed. According to the logbook and loading records for ZPR-3/6F

Lell, R. M.; McKnight, R. D; Schaefer, R. W.; Nuclear Engineering Division

2010-09-30T23:59:59.000Z

74

Hydrologic transport of depleted uranium associated with open air dynamic range testing at Los Alamos National Laboratory, New Mexico, and Eglin Air Force Base, Florida  

SciTech Connect (OSTI)

Hydrologic investigations on depleted uranium fate and transport associated with dynamic testing activities were instituted in the 1980`s at Los Alamos National Laboratory and Eglin Air Force Base. At Los Alamos, extensive field watershed investigations of soil, sediment, and especially runoff water were conducted. Eglin conducted field investigations and runoff studies similar to those at Los Alamos at former and active test ranges. Laboratory experiments complemented the field investigations at both installations. Mass balance calculations were performed to quantify the mass of expended uranium which had transported away from firing sites. At Los Alamos, it is estimated that more than 90 percent of the uranium still remains in close proximity to firing sites, which has been corroborated by independent calculations. At Eglin, we estimate that 90 to 95 percent of the uranium remains at test ranges. These data demonstrate that uranium moves slowly via surface water, in both semi-arid (Los Alamos) and humid (Eglin) environments.

Becker, N.M. [Los Alamos National Lab., NM (United States); Vanta, E.B. [Wright Laboratory Armament Directorate, Eglin Air Force Base, FL (United States)

1995-05-01T23:59:59.000Z

75

A comparison of delayed radiobiological effects of depleted-uranium munitions versus fourth-generation nuclear weapons  

E-Print Network [OSTI]

It is shown that the radiological burden due to the battle-field use of circa 400 tons of depleted-uranium munitions in Iraq (and of about 40 tons in Yugoslavia) is comparable to that arising from the hypothetical battle-field use of more than 600 kt (respectively 60 kt) of high-explosive equivalent pure-fusion fourth-generation nuclear weapons. Despite the limited knowledge openly available on existing and future nuclear weapons, there is sufficient published information on their physical principles and radiological effects to make such a comparison. In fact, it is shown that this comparison can be made with very simple and convincing arguments so that the main technical conclusions of the paper are undisputable -- although it would be worthwhile to supplement the hand calculations presented in the paper by more detailed computer simulations in order to consolidate the conclusions and refute any possible objections.

Gsponer, A; Vitale, B; Gsponer, Andre; Hurni, Jean-Pierre; Vitale, Bruno

2002-01-01T23:59:59.000Z

76

Characterization of options and their analysis requirements for the long-term management of depleted uranium hexafluoride  

SciTech Connect (OSTI)

The Department of Energy (DOE) is examining alternative strategies for the long-term management of depleted uranium hexafluoride (UF{sub 6}) currently stored at the gaseous diffusion plants at Portsmouth, Ohio, and Paducah, Kentucky, and on the Oak Ridge Reservation in Oak Ridge, Tennessee. This paper describes the methodology for the comprehensive and ongoing technical analysis of the options being considered. An overview of these options, along with several of the suboptions being considered, is presented. The long-term management strategy alternatives fall into three broad categories: use, storage, or disposal. Conversion of the depleted UF6 to another form such as oxide or metal is needed to implement most of these alternatives. Likewise, transportation of materials is an integral part of constructing the complete pathway between the current storage condition and ultimate disposition. The analysis of options includes development of pre-conceptual designs; estimates of effluents, wastes, and emissions; specification of resource requirements; and preliminary hazards assessments. The results of this analysis will assist DOE in selecting a strategy by providing the engineering information necessary to evaluate the environmental impacts and costs of implementing the management strategy alternatives.

Dubrin, J.W.; Rosen, R.S.; Zoller, J.N.; Harri, J.W.; Schwertz, N.L.

1995-12-01T23:59:59.000Z

77

Draft Environmental Impact Statement for Construction and Operation of a Depleted Uranium Hexafluoride Conversion Facility at the Portsmouth, Ohio, Site  

SciTech Connect (OSTI)

This document is a site-specific environmental impact statement (EIS) for construction and operation of a proposed depleted uranium hexafluoride (DUF{sub 6}) conversion facility at the U.S. Department of Energy (DOE) Portsmouth site in Ohio (Figure S-1). The proposed facility would convert the DUF{sub 6} stored at Portsmouth to a more stable chemical form suitable for use or disposal. The facility would also convert the DUF{sub 6} from the East Tennessee Technology Park (ETTP) site near Oak Ridge, Tennessee. In a Notice of Intent (NOI) published in the Federal Register on September 18, 2001 (Federal Register, Volume 66, page 48123 [66 FR 48123]), DOE announced its intention to prepare a single EIS for a proposal to construct, operate, maintain, and decontaminate and decommission two DUF{sub 6} conversion facilities at Portsmouth, Ohio, and Paducah, Kentucky, in accordance with the National Environmental Policy Act of 1969 (NEPA) (United States Code, Title 42, Section 4321 et seq. [42 USC 4321 et seq.]) and DOE's NEPA implementing procedures (Code of Federal Regulations, Title 10, Part 1021 [10 CFR Part 1021]). Subsequent to award of a contract to Uranium Disposition Services, LLC (hereafter referred to as UDS), Oak Ridge, Tennessee, on August 29, 2002, for design, construction, and operation of DUF{sub 6} conversion facilities at Portsmouth and Paducah, DOE reevaluated its approach to the NEPA process and decided to prepare separate site-specific EISs. This change was announced in a Federal Register Notice of Change in NEPA Compliance Approach published on April 28, 2003 (68 FR 22368); the Notice is included as Attachment B to Appendix C of this EIS. This EIS addresses the potential environmental impacts from the construction, operation, maintenance, and decontamination and decommissioning (D&D) of the proposed conversion facility at three alternative locations within the Portsmouth site; from the transportation of all ETTP cylinders (DUF{sub 6}, low-enriched UF6 [LEU-UF{sub 6}], and empty) to Portsmouth; from the transportation of depleted uranium conversion products to a disposal facility; and from the transportation, sale, use, or disposal of the fluoride-containing conversion products (hydrogen fluoride [HF] or calcium fluoride [CaF{sub 2}]). An option of shipping the ETTP cylinders to Paducah is also considered. In addition, this EIS evaluates a no action alternative, which assumes continued storage of DUF{sub 6} in cylinders at the Portsmouth and ETTP sites. A separate EIS (DOE/EIS-0359) evaluates potential environmental impacts for the proposed Paducah conversion facility.

N /A

2003-11-28T23:59:59.000Z

78

Draft Environmental Impact Statement for Construction and Operation of a Depleted Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky, Site  

SciTech Connect (OSTI)

This document is a site-specific environmental impact statement (EIS) for construction and operation of a proposed depleted uranium hexafluoride (DUF{sub 6}) conversion facility at the U.S. Department of Energy (DOE) Paducah site in northwestern Kentucky (Figure S-1). The proposed facility would convert the DUF{sub 6} stored at Paducah to a more stable chemical form suitable for use or disposal. In a Notice of Intent (NOI) published in the ''Federal Register'' (FR) on September 18, 2001 (''Federal Register'', Volume 66, page 48123 [66 FR 48123]), DOE announced its intention to prepare a single EIS for a proposal to construct, operate, maintain, and decontaminate and decommission two DUF{sub 6} conversion facilities at Portsmouth, Ohio, and Paducah, Kentucky, in accordance with the National Environmental Policy Act of 1969 (NEPA) (''United States Code'', Title 42, Section 4321 et seq. [42 USC 4321 et seq.]) and DOE's NEPA implementing procedures (''Code of Federal Regulations'', Title 10, Part 1021 [10 CFR Part 1021]). Subsequent to award of a contract to Uranium Disposition Services, LLC (hereafter referred to as UDS), Oak Ridge, Tennessee, on August 29, 2002, for design, construction, and operation of DUF{sub 6} conversion facilities at Portsmouth and Paducah, DOE reevaluated its approach to the NEPA process and decided to prepare separate site-specific EISs. This change was announced in a ''Federal Register'' Notice of Change in NEPA Compliance Approach published on April 28, 2003 (68 FR 22368); the Notice is included as Attachment B to Appendix C of this EIS. This EIS addresses the potential environmental impacts from the construction, operation, maintenance, and decontamination and decommissioning (D&D) of the proposed conversion facility at three alternative locations within the Paducah site; from the transportation of depleted uranium conversion products to a disposal facility; and from the transportation, sale, use, or disposal of the fluoride-containing conversion products (hydrogen fluoride [HF] or calcium fluoride [CaF{sub 2}]). Although not part of the proposed action, an option of shipping all cylinders (DUF{sub 6}, low-enriched UF{sub 6} [LEU-UF{sub 6}], and empty) stored at the East Tennessee Technology Park (ETTP) near Oak Ridge, Tennessee, to Paducah rather than to Portsmouth is also considered. In addition, this EIS evaluates a no action alternative, which assumes continued storage of DUF{sub 6} in cylinders at the Paducah site. A separate EIS (DOE/EIS-0360) evaluates the potential environmental impacts for the proposed Portsmouth conversion facility.

N /A

2003-11-28T23:59:59.000Z

79

Feasibility study on consolidation of Fernald Environmental Management Project depleted uranium materials  

SciTech Connect (OSTI)

In 1991, the DOE made a decision to close the FMPC located in Fernald, Ohio, and end its production mission. The site was renamed FEMP to reflect Fernald`s mission change from uranium production to environmental restoration. As a result of this change, the inventory of strategic uranium materials maintained at Fernald by DOE DP will need to be relocated to other DOE sites. Although considered a liability to the Fernald Plant due to its current D and D mission, the FEMP DU represents a potentially valuable DOE resource. Recognizing its value, it may be important for the DOE to consolidate the material at one site and place it in a safe long-term storage condition until a future DOE programmatic requirement materializes. In August 1995, the DOE Office of Nuclear Weapons Management requested, Lockheed Martin Energy Systems (LMES) to assess the feasibility of consolidating the FEMP DU materials at the Oak Ridge Reservation (ORR). This feasibility study examines various phases associated with the consolidation of the FEMP DU at the ORR. If useful short-term applications for the DU fail to materialize, then long-term storage (up to 50 years) would need to be provided. Phases examined in this report include DU material value; potential uses; sampling; packaging and transportation; material control and accountability; environmental, health and safety issues; storage; project management; noneconomic factors; schedule; and cost.

NONE

1995-11-30T23:59:59.000Z

80

ZPR-3 Assembly 12 : A cylindrical assembly of highly enriched uranium, depleted uranium and graphite with an average {sup 235}U enrichment of 21 atom %.  

SciTech Connect (OSTI)

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 12 (ZPR-3/12) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 21 at.%. Approximately 68.9% of the total fissions in this assembly occur above 100 keV, approximately 31.1% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 9 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 12 began in late Jan. 1958, and the Assembly 12 program ended in Feb. 1958. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates and graphite plates loaded into stainless steel drawers which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, seven columns of 0.125 in.-wide depleted uranium plates and seven columns of 0.125 in.-wide graphite plates. The length of each column was 9 in. (228.6 mm) in each half of the core. The graphite plates were included to produce a softer neutron spectrum that would be more characteristic of a large power reactor. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the radial blanket was approximately 12 in. and the length of the radial blanket in each half of the matrix was 21 in. (533.4 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/12, the reference critical configuration was loading 10 which was critical on Feb. 5, 1958. The subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/12 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. An accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/12 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must d

Lell, R. M.; McKnight, R. D.; Perel, R. L.; Wagschal, J. J.; Nuclear Engineering Division; Racah Inst. of Physics

2010-09-30T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Bacterial Community Succession During in situ Uranium Bioremediation: Spatial Similarities Along Controlled Flow Paths  

E-Print Network [OSTI]

problem, and the use of depleted uranium and other heavyenvironmental hazard. Depleted uranium is weakly radioactiveMB. (2004). Depleted and natural uranium: chemistry and

Hwang, Chiachi

2009-01-01T23:59:59.000Z

82

A novel hohlraum with ultrathin depleted-uranium-nitride coating layer for low hard x-ray emission and high radiation temperature  

E-Print Network [OSTI]

An ultra-thin layer of uranium nitrides (UN) has been coated on the inner surface of the depleted uranium hohlraum (DUH), which has been proved by our experiment can prevent the oxidization of Uranium (U) effectively. Comparative experiments between the novel depleted uranium hohlraum and pure golden (Au) hohlraum are implemented on Shenguang III prototype laser facility. Under the laser intensity of 6*10^14 W/cm2, we observe that, the hard x-ray (> 1.8 keV) fraction of this uranium hohlraum decreases by 61% and the peak intensity of total x-ray flux (0.1 keV ~ 5 keV) increases by 5%. Two dimensional radiation hydrodynamic code LARED are exploited to interpret the above observations. Our result for the first time indicates the advantage of the UN-coated DUH in generating the uniform x-ray field with a quasi Planckian spectrum and thus has important implications in optimizing the ignition hohlraum design.

Guo, Liang; Xing, Peifeng; Li, Sanwei; Yi, Taimin; Kuang, Longyu; Li, Zhichao; Li, Renguo; Wu, Zheqing; Jing, Longfei; Zhang, Wenhai; Zhan, Xiayu; Yang, Dong; Jiang, Bobi; Yang, Jiamin; Liu, Shenye; Jiang, Shaoen; Li, Yongsheng; Liu, Jie; Huo, Wenyi; Lan, Ke

2014-01-01T23:59:59.000Z

83

Long-term fate of depleted uranium at Aberdeen and Yuma Proving Grounds: Human health and ecological risk assessments  

SciTech Connect (OSTI)

The purpose of this study was to evaluate the immediate and long-term consequences of depleted uranium (DU) in the environment at Aberdeen Proving Ground (APG) and Yuma Proving Ground (YPG) for the Test and Evaluation Command (TECOM) of the US Army. Specifically, we examined the potential for adverse radiological and toxicological effects to humans and ecosystems caused by exposure to DU at both installations. We developed contaminant transport models of aquatic and terrestrial ecosystems at APG and terrestrial ecosystems at YPG to assess potential adverse effects from DU exposure. Sensitivity and uncertainty analyses of the initial models showed the portions of the models that most influenced predicted DU concentrations, and the results of the sensitivity analyses were fundamental tools in designing field sampling campaigns at both installations. Results of uranium (U) isotope analyses of field samples provided data to evaluate the source of U in the environment and the toxicological and radiological doses to different ecosystem components and to humans. Probabilistic doses were estimated from the field data, and DU was identified in several components of the food chain at APG and YPG. Dose estimates from APG data indicated that U or DU uptake was insufficient to cause adverse toxicological or radiological effects. Dose estimates from YPG data indicated that U or DU uptake is insufficient to cause radiological effects in ecosystem components or in humans, but toxicological effects in small mammals (e.g., kangaroo rats and pocket mice) may occur from U or DU ingestion. The results of this study were used to modify environmental radiation monitoring plans at APG and YPG to ensure collection of adequate data for ongoing ecological and human health risk assessments.

Ebinger, M.H.; Beckman, R.J.; Myers, O.B. [Los Alamos National Lab., NM (United States); Kennedy, P.L.; Clements, W.; Bestgen, H.T. [Colorado State Univ., Ft. Collins, CO (United States). Dept. of Fishery and Wildlife Biology

1996-09-01T23:59:59.000Z

84

acute tryptophan depletion: Topics by E-print Network  

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

for Psychopharmacology ISSN 0269-8811 SAGE Publications Ltd 12 Review The Toxicity of Depleted Uranium CiteSeer Summary: Abstract: Depleted uranium (DU) is an emerging...

85

Depleted uranium risk assessment for Jefferson Proving Ground using data from environmental monitoring and site characterization. Final report  

SciTech Connect (OSTI)

This report documents the third risk assessment completed for the depleted uranium (DU) munitions testing range at Jefferson Proving Ground (JPG), Indiana, for the U.S. Army Test and Evaluation command. Jefferson Proving Ground was closed in 1995 under the Base Realignment and Closure Act and the testing mission was moved to Yuma Proving Ground. As part of the closure of JPG, assessments of potential adverse health effects to humans and the ecosystem were conducted. This report integrates recent information obtained from site characterization surveys at JPG with environmental monitoring data collected from 1983 through 1994 during DU testing. Three exposure scenarios were evaluated for potential adverse effects to human health: an occasional use scenario and two farming scenarios. Human exposure was minimal from occasional use, but significant risk were predicted from the farming scenarios when contaminated groundwater was used by site occupants. The human health risk assessments do not consider the significant risk posed by accidents with unexploded ordnance. Exposures of white-tailed deer to DU were also estimated in this study, and exposure rates result in no significant increase in either toxicological or radiological risks. The results of this study indicate that remediation of the DU impact area would not substantially reduce already low risks to humans and the ecosystem, and that managed access to JPG is a reasonable model for future land use options.

Ebinger, M.H.; Hansen, W.R.

1996-10-01T23:59:59.000Z

86

Radiation- and Depleted Uranium-Induced Carcinogenesis Studies: Characterization of the Carcinogenic Process and Development of Medical Countermeasures  

E-Print Network [OSTI]

External or internal contamination from radioactive elements during military operations or a terrorist attack is a serious threat to military and civilian populations. External radiation exposure could result from conventional military scenarios including nuclear weapons use and low-dose exposures during radiation accidents or terrorist attacks. Alternatively, internal radiation exposure could result from depleted uranium exposure via DU shrapnel wounds or inhalation. The long-term health effects of these types of radiation exposures are not well known. Furthermore, development of pharmacological countermeasures to low-dose external and internal radiological contamination is essential to the health and safety of both military and civilian populations. The purpose of these studies is to evaluate low-dose radiation or DU-induced carcinogenesis using in vitro and in vivo models, and to test safe and efficacious medical countermeasures. A third goal of these studies is to identify biomarkers of both exposure and disease development. Initially, we used a human cell model (human osteoblast cells, HOS) to evaluate the carcinogenic potential of DU in vitro by assessing morphological transformation, genotoxicity (chromosomal aberrations), mutagenic (HPRT loci), and genomic instability. As a comparison, low-dose cobalt radiation, broad-beam alpha particles, and other military-projectile metals, i.e., tungsten mixtures, are being examined. Published data from

A. C. Miller; D. Beltran; R. Rivas; M. Stewart; R. J. Merlot; P. B. Lison

87

Radiation shielding materials and containers incorporating same  

DOE Patents [OSTI]

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound ("PYRUC") shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Mirsky, Steven M. (Greenbelt, MD); Krill, Stephen J. (Arlington, VA); Murray, Alexander P. (Gaithersburg, MD)

2005-11-01T23:59:59.000Z

88

Radiation Shielding Materials and Containers Incorporating Same  

DOE Patents [OSTI]

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

2005-11-01T23:59:59.000Z

89

Control of structure and reactivity by ligand design : applications to small molecule activation by low-valent uranium complexes  

E-Print Network [OSTI]

coordination chemistry is depleted uranium, a by-product innuclear reactors. Depleted uranium Figure 1-1. The periodic

Lam, Oanh Phi

2010-01-01T23:59:59.000Z

90

Streamlined approach for environmental restoration plan for corrective action unit 430, buried depleted uranium artillery round No. 1, Tonopah test range  

SciTech Connect (OSTI)

This plan addresses actions necessary for the restoration and closure of Corrective Action Unit (CAU) No. 430, Buried Depleted Uranium (DU) Artillery Round No. 1 (Corrective Action Site No. TA-55-003-0960), a buried and unexploded W-79 Joint Test Assembly (JTA) artillery test projectile with high explosives (HE), at the U.S. Department of Energy, Nevada Operations Office (DOE/NV) Tonopah Test Range (TTR) in south-central Nevada. It describes activities that will occur at the site as well as the steps that will be taken to gather adequate data to obtain a notice of completion from Nevada Division of Environmental Protection (NDEP). This plan was prepared under the Streamlined Approach for Environmental Restoration (SAFER) concept, and it will be implemented in accordance with the Federal Facility Agreement and Consent Order (FFACO) and the Resource Conservation and Recovery Act (RCRA) Industrial Sites Quality Assurance Project Plan.

NONE

1996-09-01T23:59:59.000Z

91

Biological assessment of the effects of construction and operation of a depleted uranium hexafluoride conversion facility at the Paducah, Kentucky, site.  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) Depleted Uranium Hexafluoride (DUF{sub 6}) Management Program evaluated alternatives for managing its inventory of DUF{sub 6} and issued the ''Programmatic Environmental Impact Statement for Alternative Strategies for the Long-Term Management and Use of Depleted Uranium Hexafluoride'' (DUF{sub 6} PEIS) in April 1999 (DOE 1999). The DUF{sub 6} inventory is stored in cylinders at three DOE sites: Paducah, Kentucky; Portsmouth, Ohio; and East Tennessee Technology Park (ETTP), near Oak Ridge, Tennessee. In the Record of Decision for the DUF{sub 6} PEIS, DOE stated its decision to promptly convert the DUF6 inventory to a more stable chemical form. Subsequently, the U.S. Congress passed, and the President signed, the ''2002 Supplemental Appropriations Act for Further Recovery from and Response to Terrorist Attacks on the United States'' (Public Law No. 107-206). This law stipulated in part that, within 30 days of enactment, DOE must award a contract for the design, construction, and operation of a DUF{sub 6} conversion plant at the Department's Paducah, Kentucky, and Portsmouth, Ohio, sites, and for the shipment of DUF{sub 6} cylinders stored at ETTP to the Portsmouth site for conversion. This biological assessment (BA) has been prepared by DOE, pursuant to the National Environmental Policy Act of 1969 (NEPA) and the Endangered Species Act of 1974, to evaluate potential impacts to federally listed species from the construction and operation of a conversion facility at the DOE Paducah site.

Van Lonkhuyzen, R.

2005-09-09T23:59:59.000Z

92

acute catecholamine depletion: Topics by E-print Network  

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

about NM biosynthesis, and it is not known where Sulzer, David 9 Review The Toxicity of Depleted Uranium CiteSeer Summary: Abstract: Depleted uranium (DU) is an emerging...

93

Overview of the Capstone Depleted Uranium Study of Aerosols from Impact with Armored Vehicles: Test Setup and Aerosol Generation, Characterization, and Application in Assessing Dose and Risk  

SciTech Connect (OSTI)

The Capstone Depleted Uranium (DU) Aerosol Characterization and Risk Assessment Study was conducted to generate data about DU aerosols generated during the perforation of armored combat vehicles with large-caliber DU penetrators, and to apply the data in assessments of human health risks to personnel exposed to these aerosols, primarily through inhalation, during the 1991 Gulf War or in future military operations. The Capstone study consisted of two components: 1) generating, sampling and characterizing DU aerosols by firing at and perforating combat vehicles and 2) applying the source-term quantities and characteristics of the aerosols to the evaluation of doses and risks. This paper reviews the background of the study including the bases for the study, previous reviews of DU particles and health assessments from DU used by the U.S. military, the objectives of the study components, the participants and oversight teams, and the types of exposures it was intended to evaluate. It then discusses exposure scenarios used in the dose and risk assessment and provides an overview of how the field tests and dose and risk assessments were conducted.

Parkhurst, MaryAnn; Guilmette, Raymond A.

2009-03-01T23:59:59.000Z

94

Radiation shielding composition  

DOE Patents [OSTI]

A composition for use as a radiation shield. The shield is a concrete product containing a stable uranium aggregate for attenuating gamma rays and a neutron absorbing component, the uranium aggregate and neutron absorbing component being present in the concrete product in sufficient amounts to provide a concrete having a density between about 4 and about 15 grams/cm.sup.3 and which will at a predetermined thickness, attenuate gamma rays and absorb neutrons from a radioactive material of projected gamma ray and neutron emissions over a determined time period. The composition is preferably in the form of a container for storing radioactive materials that emit gamma rays and neutrons. The concrete container preferably comprises a metal liner and/or a metal outer shell. The resulting radiation shielding container has the potential of being structurally sound, stable over a long period of time, and, if desired, readily mobile.

Quapp, William J. (Idaho Falls, ID); Lessing, Paul A. (Idaho Falls, ID)

2000-12-26T23:59:59.000Z

95

Radiation shielding composition  

DOE Patents [OSTI]

A composition for use as a radiation shield. The shield is a concrete product containing a stable uranium aggregate for attenuating gamma rays and a neutron absorbing component, the uranium aggregate and neutron absorbing component being present in the concrete product in sufficient amounts to provide a concrete having a density between about 4 and about 15 grams/cm.sup.3 and which will at a predetermined thickness, attenuate gamma rays and absorb neutrons from a radioactive material of projected gamma ray and neutron emissions over a determined time period. The composition is preferably in the form of a container for storing radioactive materials that emit gamma rays and neutrons. The concrete container preferably comprises a metal liner and/or a metal outer shell. The resulting radiation shielding container has the potential of being structurally sound, stable over a long period of time, and, if desired, readily mobile.

Quapp, William J. (Idaho Falls, ID); Lessing, Paul A. (Idaho Falls, ID)

1998-01-01T23:59:59.000Z

96

Radiation shielding composition  

DOE Patents [OSTI]

A composition is disclosed for use as a radiation shield. The shield is a concrete product containing a stable uranium aggregate for attenuating gamma rays and a neutron absorbing component, the uranium aggregate and neutron absorbing component being present in the concrete product in sufficient amounts to provide a concrete having a density between about 4 and about 15 grams/cm{sup 3} and which will at a predetermined thickness, attenuate gamma rays and absorb neutrons from a radioactive material of projected gamma ray and neutron emissions over a determined time period. The composition is preferably in the form of a container for storing radioactive materials that emit gamma rays and neutrons. The concrete container preferably comprises a metal liner and/or a metal outer shell. The resulting radiation shielding container has the potential of being structurally sound, stable over a long period of time, and, if desired, readily mobile. 5 figs.

Quapp, W.J.; Lessing, P.A.

1998-07-28T23:59:59.000Z

97

WISE Uranium Project - Fact Sheet  

E-Print Network [OSTI]

t in the depleted uranium. For this purpose, we first need to calculate the mass balance of the enrichment process. We then calculate the inhalation doses from the depleted uranium and compare the dose contributions from the nuclides of interest. Mass balance for uranium enrichment at Paducah [DOE_1984, p.35] Feed Product Tails Other Mass [st] 758002 124718 621894 11390 Mass fraction 100.00% 16.45% 82.04% 1.50% Concentration of plutonium in tails (depleted uranium) from enrichment of reprocessed uranium, assuming that all plutonium were transfered to the tails: Concentration of neptunium in tails from enrichment of reprocessed uranium uranium, assuming that all neptunium were transfered to the tails: - 2 - Schematic of historic uranium enrichment process at Paducah [DOE_1999b] - -7 For comparison, we first calculate the inhalation dose from depleted uranium produced from natural uranium. We assume that the short-lived decay products have reached secular equilibrium with th

Hazards From Depleted

98

Magnetic Exchange Coupling and Single-Molecule Magnetism in Uranium Complexes  

E-Print Network [OSTI]

greater than 99% U-238 (depleted uranium), which has no neturanium, since this actinide element offers minimal radioactivity (in depleted

Rinehart, Jeffrey Dennis

2010-01-01T23:59:59.000Z

99

Controlling uranium reactivity March 18, 2008  

E-Print Network [OSTI]

for the last decade. Most of their work involves depleted uranium, a more common form of uraniumMarch 2008 Controlling uranium reactivity March 18, 2008 Uranium is an often misunderstood metal uranium research. In reality, uranium presents a wealth of possibilities for funda- mental chemistry. Many

Meyer, Karsten

100

Safe Operating Procedure SAFETY PROTOCOL: URANIUM  

E-Print Network [OSTI]

involve the use of natural or depleted uranium. Natural isotopes of uranium are U-238, U-235 and U-234 (see Table 1 for natural abundances). Depleted uranium contains less of the isotopes: U-235 and U-234. The specific activity of depleted uranium (5.0E-7 Ci/g) is less than that of natural uranium (7.1E-7 Ci

Farritor, Shane

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications  

E-Print Network [OSTI]

The sintering behavior of uranium and uranium-zirconium alloys in the alpha phase were characterized in this research. Metal uranium powder was produced from pieces of depleted uranium metal acquired from the Y-12 plant via hydriding...

Helmreich, Grant

2012-02-14T23:59:59.000Z

102

Disposition of Depleted Uranium Oxide  

SciTech Connect (OSTI)

This document summarizes environmental information which has been collected up to June 1983 at Savannah River Plant. Of particular interest is an updating of dose estimates from changes in methodology of calculation, lower cesium transport estimates from Steel Creek, and new sports fish consumption data for the Savannah River. The status of various permitting requirements are also discussed.

Crandall, J.L.

2001-08-13T23:59:59.000Z

103

URANIUM MILLING ACTIVITIES AT SEQUOYAH FUELS CORPORATION  

E-Print Network [OSTI]

Sequoyah Fuels Corporation (SFC) describes previous operations at its Gore, Oklahoma, uranium conversion facility as: (1) the recovery of uranium by concentration and purification processes; and (2) the conversion of concentrated and purified uranium ore into uranium hexafluoride (UF 6), or the reduction of depleted uranium tetrafluoride (UF 4) to UF 6. SFC contends that these

unknown authors

104

atp depletion precedes: Topics by E-print Network  

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

evolves, because new precedents are generated by the form... Smolin, Lee 2012-01-01 15 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

105

analogues deplete androgen: Topics by E-print Network  

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

with androgens has been shown to increase growth rate in fishes (Ron et al., 1995 13 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

106

antioxidant defence depletion: Topics by E-print Network  

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

and defence reactions. Priya Roy; Ramamurthy Dhandapani Department Of Microbiology 15 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

107

administration depletes mitochondrial: Topics by E-print Network  

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

recombination is widespread in plant mtDNA. Recombinant molecules have Nicolas Galtier 6 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

108

approaching waterflood depletion: Topics by E-print Network  

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

are shown in Table 5 of the Appendix. Figure... Pettitt, Bobby Eugene 1963-01-01 19 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

109

DOE Selects Contractor for Depleted Hexafluoride Conversion Project...  

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

to the DOE Portsmouth Paducah Project Office (PPPO) in Lexington, Kentucky and the Depleted Uranium Hexafluoride (DUF6) Conversion Project in Paducah, Kentucky and...

110

aerosol depletion test: Topics by E-print Network  

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

damage and realize optimum well productivity. To address... Chen, Guoqiang 2002-01-01 10 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

111

Uranium Acquisition | Y-12 National Security Complex  

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

of Interest (EOI) to acquire up to 6,800 metric tons of Uranium (MTU) of high purity depleted uranium metal (DU) and related material and services. This request for EOI does...

112

Welding of uranium and uranium alloys  

SciTech Connect (OSTI)

The major reported work on joining uranium comes from the USA, Great Britain, France and the USSR. The driving force for producing this technology base stems from the uses of uranium as a nuclear fuel for energy production, compact structures requiring high density, projectiles, radiation shielding, and nuclear weapons. This review examines the state-of-the-art of this technology and presents current welding process and parameter information. The welding metallurgy of uranium and the influence of microstructure on mechanical properties is developed for a number of the more commonly used welding processes.

Mara, G.L.; Murphy, J.L.

1982-03-26T23:59:59.000Z

113

Thermocouple shield  

DOE Patents [OSTI]

A thermocouple shield for use in radio frequency fields. In some embodiments the shield includes an electrically conductive tube that houses a standard thermocouple having a thermocouple junction. The electrically conductive tube protects the thermocouple from damage by an RF (including microwave) field and mitigates erroneous temperature readings due to the microwave or RF field. The thermocouple may be surrounded by a ceramic sheath to further protect the thermocouple. The ceramic sheath is generally formed from a material that is transparent to the wavelength of the microwave or RF energy. The microwave transparency property precludes heating of the ceramic sheath due to microwave coupling, which could affect the accuracy of temperature measurements. The ceramic sheath material is typically an electrically insulating material. The electrically insulative properties of the ceramic sheath help avert electrical arcing, which could damage the thermocouple junction. The electrically conductive tube is generally disposed around the thermocouple junction and disposed around at least a portion of the ceramic sheath. The concepts of the thermocouple shield may be incorporated into an integrated shielded thermocouple assembly.

Ripley, Edward B. (Knoxville, TN)

2009-11-24T23:59:59.000Z

114

activity-dependent vmat-mediated depletion: Topics by E-print...  

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

by which this main- tenance is achieved. Its functions include Huettner, James E. 3 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

115

ampt-induced monoamine depletion: Topics by E-print Network  

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

monoaminergic and peptidergic signaling due (more) Wragg, Rachel T. 2010-01-01 8 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

116

acid depleted space-flown: Topics by E-print Network  

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

D Hermier 1, D Catheline 2,D Hermier D Catheline Paris-Sud XI, Universit de 2 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

117

androgen depletion up-regulates: Topics by E-print Network  

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

with androgens has been shown to increase growth rate in fishes (Ron et al., 1995 17 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

118

Pulsed, Photonuclear-induced, Neutron Measurements of Nuclear Materials with Composite Shielding  

SciTech Connect (OSTI)

Active measurements were performed using a 10-MeV electron accelerator with inspection objects containing various nuclear and nonnuclear materials available at the Idaho National Laboratory’s Zero Power Physics Reactor (ZPPR) facility. The inspection objects were assembled from ZPPR reactor plate materials to evaluate the measurement technologies for the characterization of plutonium, depleted uranium or highly enriched uranium shielded by both nuclear and non-nuclear materials. A series of pulsed photonuclear, time-correlated measurements were performed with unshielded calibration materials and then compared with the more complex composite shield configurations. The measurements used multiple 3He detectors that are designed to detect fission neutrons between pulses of an electron linear accelerator. The accelerator produced 10-MeV bremsstrahlung X-rays at a repetition rate of 125 Hz (8 ms between pulses) with a 4-us pulse width. All inspected objects were positioned on beam centerline and 100 cm from the X-ray source. The time-correlated data was collected in parallel using both a Los Alamos National Laboratory-designed list-mode acquisition system and a commercial multichannel scaler analyzer. A combination of different measurement configurations and data analysis methods enabled the identification of each object. This paper describes the experimental configuration, the ZPPR inspection objects used, and the various measurement and analysis results for each inspected object.

James Jones; Kevin Haskell; Rich Waston; William Geist; Jonathan Thron; Corey Freeman; Martyn Swinhoe; Seth McConchie; Eric Sword; Lee Montierth; John Zabriskie

2011-07-01T23:59:59.000Z

119

alarming oxygen depletion: Topics by E-print Network  

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

T. Doan; Q. Cao; L. Selavo; Y. Wu; L. Fang; Z. He; S. Lin; J. Stankovic 2006-01-01 37 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

120

Radiochemical Analysis Methodology for uranium Depletion Measurements  

SciTech Connect (OSTI)

This report provides sufficient material for a test sponsor with little or no radiochemistry background to understand and follow physics irradiation test program execution. Most irradiation test programs employ similar techniques and the general details provided here can be applied to the analysis of other irradiated sample types. Aspects of program management directly affecting analysis quality are also provided. This report is not an in-depth treatise on the vast field of radiochemical analysis techniques and related topics such as quality control. Instrumental technology is a very fast growing field and dramatic improvements are made each year, thus the instrumentation described in this report is no longer cutting edge technology. Much of the background material is still applicable and useful for the analysis of older experiments and also for subcontractors who still retain the older instrumentation.

Scatena-Wachel DE

2007-01-09T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Magnetic shielding  

DOE Patents [OSTI]

A magnetically-conductive filler material bridges the gap between a multi-part magnetic shield structure which substantially encloses a predetermined volume so as to minimize the ingress or egress of magnetic fields with respect to that volume. The filler material includes a heavy concentration of single-magnetic-domain-sized particles of a magnetically conductive material (e.g. soft iron, carbon steel or the like) dispersed throughout a carrier material which is generally a non-magnetic material that is at least sometimes in a plastic or liquid state. The maximum cross-sectional particle dimension is substantially less than the nominal dimension of the gap to be filled. An epoxy base material (i.e. without any hardening additive) low volatility vacuum greases or the like may be used for the carrier material. The structure is preferably exposed to the expected ambient field while the carrier is in a plastic or liquid state so as to facilitate alignment of the single-magnetic-domain-sized particles with the expected magnetic field lines.

Kerns, J.A.; Stone, R.R.; Fabyan, J.

1985-02-12T23:59:59.000Z

122

Disposition of DOE Excess Depleted Uranium, Natural Uranium, and  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeat Pump Models |Conduct, Parent(CRADA and DOW Area 5 LLRW & MLLWLow-Enriched

123

Detection of uranium-based nuclear weapons using neutron-induced fission  

SciTech Connect (OSTI)

Although plutonium-based nuclear weapons can usually be detected by their spontaneous emission of neutrons and gammas, the radiation emitted by weapons based entirely on highly-enriched uranium can often be easily shielded. Verification of a treaty that limits the number of such weapons may require an active technique, such as interrogating the suspect assembly with an external neutron source and measuring the number of fission neutrons produced. Difficulties include distinguishing between source and fission neutrons, the variations in yield for different materials and geometries, and the possibility of non-nuclear weapons that may contain significant amounts of fissionable depleted uranium. We describe simple measurements that test the induced-fission technique using an isotopic Am-Li source, an novel energy-sensitive neutron detector, and several small assemblies containing {sup 235}U, {sup 238}U, lead, and polyethylene. In all cases studied, the neutron yields above the source energy are larger for the {sup 235}U assemblies than for assemblies containing only lead or depleted uranium. For more complex geometries, corrections for source transmission may be necessary. The results are promising enough to recommend further experiments and calculations using examples of realistic nuclear and non-nuclear weapons. 5 refs., 11 figs.

Moss, C.E.; Byrd, R.C.; Feldman, W.C.; Auchampaugh, G.F.; Estes, G.P. [Los Alamos National Lab., NM (United States); Ewing, R.I.; Marlow, K.W. [Sandia National Labs., Albuquerque, NM (United States)

1991-12-01T23:59:59.000Z

124

Detection of uranium-based nuclear weapons using neutron-induced fission  

SciTech Connect (OSTI)

Although plutonium-based nuclear weapons can usually be detected by their spontaneous emission of neutrons and gammas, the radiation emitted by weapons based entirely on highly-enriched uranium can often be easily shielded. Verification of a treaty that limits the number of such weapons may require an active technique, such as interrogating the suspect assembly with an external neutron source and measuring the number of fission neutrons produced. Difficulties include distinguishing between source and fission neutrons, the variations in yield for different materials and geometries, and the possibility of non-nuclear weapons that may contain significant amounts of fissionable depleted uranium. We describe simple measurements that test the induced-fission technique using an isotopic Am-Li source, an novel energy-sensitive neutron detector, and several small assemblies containing {sup 235}U, {sup 238}U, lead, and polyethylene. In all cases studied, the neutron yields above the source energy are larger for the {sup 235}U assemblies than for assemblies containing only lead or depleted uranium. For more complex geometries, corrections for source transmission may be necessary. The results are promising enough to recommend further experiments and calculations using examples of realistic nuclear and non-nuclear weapons. 5 refs., 11 figs.

Moss, C.E.; Byrd, R.C.; Feldman, W.C.; Auchampaugh, G.F.; Estes, G.P. (Los Alamos National Lab., NM (United States)); Ewing, R.I.; Marlow, K.W. (Sandia National Labs., Albuquerque, NM (United States))

1991-01-01T23:59:59.000Z

125

Uranium Ore Uranium is extracted  

E-Print Network [OSTI]

Milling of Uranium Ore Uranium is extracted from ore with strong acids or bases. The uranium is concentrated in a solid substance called"yellowcake." Chemical Conversion Plants convert the uranium in yellowcake to uranium hexafluoride (UF6 ), a compound that can be made into nuclear fuel. Enrichment

126

Effect of Shim Arm Depletion in the NBSR  

SciTech Connect (OSTI)

The cadmium shim arms in the NBSR undergo burnup during reactor operation and hence, require periodic replacement. Presently, the shim arms are replaced after every 25 cycles to guarantee they can maintain sufficient shutdown margin. Two prior reports document the expected change in the 113Cd distribution because of the shim arm depletion. One set of calculations was for the present high-enriched uranium fuel and the other for the low-enriched uranium fuel when it was in the COMP7 configuration (7 inch fuel length vs. the present 11 inch length). The depleted 113Cd distributions calculated for these cores were applied to the current design for an equilibrium low-enriched uranium core. This report details the predicted effects, if any, of shim arm depletion on the shim arm worth, the shutdown margin, power distributions and kinetics parameters.

Hanson A. H.; Brown N.; Diamond, D.J.

2013-02-22T23:59:59.000Z

127

Uranium Oxide as a Highly Reflective Coating from 150-350 eV  

E-Print Network [OSTI]

of depleted uranium metal (less than 0.2% U-235). After sputtering, the uranium was allowed to oxidize1 Uranium Oxide as a Highly Reflective Coating from 150-350 eV Richard L. Sandberg, David D. Allred.byu.edu ABSTRACT We present the measured reflectances (beamline 6.3.2, ALS at LBNL) of naturally oxidized uranium

Hart, Gus

128

Shielding and criticality analyses of phase I reference truck and rail cask designs for spent nuclear fuel  

SciTech Connect (OSTI)

Results are presented herein to determine the adequacy with respect to shielding regulations of reference designs for a truck cask containing 2 PWR or 5 BWR assemblies of standard burnup (45 GWd/MTU for PWR, 40 GWd/MTU for BWR) and 1 PWR assembly with extended burnup (55 GWd/MTU). The study also includes reference and modified rail cask designs with projected payloads of 8, 10, or 12 PWR assemblies. The burnup/age trends are analyzed in one dimension for both Pb and depleted uranium (DU) gamma-ray shields. The results of the two-dimensional shielding analysis uphold the one-dimensional results as being an appropriate means of studying the burnup/age trends for the truck cask. These results show that the reference design for the Pb-shield truck cask is inadequate for all cases considered, while the DU-shield truck cask is capable of carrying the desired payloads. The one-dimensional shielding analysis results for the reference Pb and DU rail casks indicate substantial margins exist in the side doses for reasonable burnup/age combinations. For a Pb-cask configuration, margins exist primarily for long-cooled (15 years) fuel. For the modified Pb and DU rail casks, the 2-m dose rates offer substantial margins below the regulatory limits for all burnup values considered provided the spent fuel has cooled for {>=}10 years. The modified Pb and DU casks yield essentially identical results and, hence, could be considered equivalent from a shielding perspective. The criticality analyses that were performed indicate that a truck basket can be designed to provide an adequate subcritical margin for 2 PWR assemblies enriched to 5 wt%. While the 10- and 12- assembly rail cask designs are very close to the regulatory limit of 0.95 for k{sub eff}, after accounting for a 0.01 {Delta}k bias and 2 standard deviations, the limit is exceeded by about 3%. It is believed that a combination of decreased enrichments and/or increased water gaps should allow these baskets to be acceptable.

Broadhead, B.L.; Childs, R.L.; Parks, C.V.

1996-03-01T23:59:59.000Z

129

Uranium deposits of Brazil  

SciTech Connect (OSTI)

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

NONE

1991-09-01T23:59:59.000Z

130

INTOR radiation shielding for personnel access  

SciTech Connect (OSTI)

The INTOR reactor shield system consists of the blanket, bulk shield, penetration shield, component shield, and biological shield. The bulk shield consists of two parts: (a) the inboard shield; and (b) the outboard shield. The distinction between the different components of the shield system is essential to satisfy the different design constraints and achieve various objectives.

Gohar, Y.; Abdou, M.

1981-01-01T23:59:59.000Z

131

Gamma ray detector shield  

DOE Patents [OSTI]

A gamma ray detector shield comprised of a rigid, lead, cylindrical-shaped vessel having upper and lower portions with an pneumatically driven, sliding top assembly. Disposed inside the lead shield is a gamma ray scintillation crystal detector. Access to the gamma detector is through the sliding top assembly.

Ohlinger, R.D.; Humphrey, H.W.

1985-08-26T23:59:59.000Z

132

arginase-induced l-arginine depletion: Topics by E-print Network  

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

in presence or absence of L-arginine. N-hydroxy-nor-l- arginine (nor-NOHA) and alpha 13 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

133

allogeneic t-cell depleted: Topics by E-print Network  

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

T cells expressed aid mRNA as well as AID protein. We Paris-Sud XI, Universit de 52 Depleted Uranium Technical Brief Environmental Sciences and Ecology Websites Summary: and...

134

Adhesive particle shielding  

DOE Patents [OSTI]

An efficient device for capturing fast moving particles has an adhesive particle shield that includes (i) a mounting panel and (ii) a film that is attached to the mounting panel wherein the outer surface of the film has an adhesive coating disposed thereon to capture particles contacting the outer surface. The shield can be employed to maintain a substantially particle free environment such as in photolithographic systems having critical surfaces, such as wafers, masks, and optics and in the tools used to make these components, that are sensitive to particle contamination. The shield can be portable to be positioned in hard-to-reach areas of a photolithography machine. The adhesive particle shield can incorporate cooling means to attract particles via the thermophoresis effect.

Klebanoff, Leonard Elliott (Dublin, CA); Rader, Daniel John (Albuquerque, NM); Walton, Christopher (Berkeley, CA); Folta, James (Livermore, CA)

2009-01-06T23:59:59.000Z

135

SHIELD certification package  

SciTech Connect (OSTI)

Certification as applied to existing computer codes includes the verification and validation process, placing the code in configuration control, establishing user qualification standards and training requirements. All software intended for use in critical calculations must be certified. This report is intended to fulfill the requirements for the certification of the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system built February, 1992, by W.S. Parks. These modules are used for burnup, cooling, separate, and edit calculations.

Boman, C.

1992-02-01T23:59:59.000Z

136

Solid State Phase Transformations in Uranium-Zirconium Alloys  

E-Print Network [OSTI]

Depleted Uranium WDS Wavelength Dispersive Spectroscopy DIC Di erential Interference Contrast BSE Back Scattered Electron image SS Stainless Steel MIC Microscopy and Imaging Center OR Orientation Relationship EDS Energy Dispersive Spectroscopy UNLV...

Irukuvarghula, Sandeep

2013-08-06T23:59:59.000Z

137

JOURNAL DE PHYSIQUE Colloque C7, supplment au n12, Tome 43, dcembre 1982 page C7-45  

E-Print Network [OSTI]

shielding block made of masonite is placed on the same carriage. Depleted uranium and lead blocks acting

Paris-Sud XI, Université de

138

Radation shielding pellets  

DOE Patents [OSTI]

Radiation pellets having an outer shell, preferably, of Mo, W or depleted U nd an inner filling of lithium hydride wherein the outer shell material has a greater melting point than does the inner filling material.

Coomes, Edmund P. (Richland, WA); Luksic, Andrzej T. (Pasco, WA)

1988-01-01T23:59:59.000Z

139

PREPARED FOR THE U.S. DEPARTMENT OF ENERGY, UNDER CONTRACT DE-AC02-76CH03073  

E-Print Network [OSTI]

and Disposal of a Radium-beryllium Source Using a Depleted Uranium Polyethylene Composite Shielding Container-BERYLLIUM SOURCE USING DEPLETED URANIUM POLYETHYLENE COMPOSITE SHIELDING Keith Rule Princeton Plasma Physics of depleted uranium oxide encapsulated in polyethylene to provide suitable shielding for both gamma

140

Enclosure 1 -CCP-AK-INL-004, Table 5-2 (1 page) Table 5-2. Isotopic Compositions of Rocky Flats Plutonium and Uranium  

E-Print Network [OSTI]

Flats Plutonium and Uranium Weapons-Grade Plutonium Enriched Uranium Depleted Uranium Plutonium-238 0.01 ­ 0.05% Uranium-234 0.1 ­ 1.02% Uranium-234 0.0006% Plutonium-239 92.8 ­ 94.4% Uranium-235 90 ­ 94% Uranium-235 0.2 ­ 0.3% Plutonium-240 4.85 ­ 6.5% Uranium-236 0.4 ­ 0.5% Uranium-238 99.7 ­ 99.8% Plutonium

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Lightweight blast shield  

DOE Patents [OSTI]

A tandem warhead missile arrangement that has a composite material housing structure with a first warhead mounted at one end and a second warhead mounted near another end of the composite structure with a dome shaped composite material blast shield mounted between the warheads to protect the second warhead from the blast of the first warhead.

Mixon, Larry C. (Madison, AL); Snyder, George W. (Huntsville, AL); Hill, Scott D. (Toney, AL); Johnson, Gregory L. (Decatur, AL); Wlodarski, J. Frank (Huntsville, AL); von Spakovsky, Alexis P. (Huntsville, AL); Emerson, John D. (Arab, AL); Cole, James M. (Huntsville, AL); Tipton, John P. (Huntsville, AL)

1991-01-01T23:59:59.000Z

142

Review Article RADIATION SHIELDING TECHNOLOGY  

E-Print Network [OSTI]

Review Article RADIATION SHIELDING TECHNOLOGY J. Kenneth Shultis and Richard E. Faw* Abstract Physics Society INTRODUCTION THIS IS a review of the technology of shielding against the effects to the review. The first treats the evolution of radiation-shielding technology from the beginning of the 20th

Shultis, J. Kenneth

143

SUPERCONDUCTING SHIELDING By W. O. HAMILTON,  

E-Print Network [OSTI]

41. SUPERCONDUCTING SHIELDING By W. O. HAMILTON, Stanford University, Department of Physics, Stanford, California (U.S.A.). Abstract. 2014 Superconducting shields offer the possibility of obtaining shielding from external time varying fields. Various techniques of superconducting shielding

Paris-Sud XI, Université de

144

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

SciTech Connect (OSTI)

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.

NONE

1995-07-05T23:59:59.000Z

145

Uranium industry annual 1997  

SciTech Connect (OSTI)

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

NONE

1998-04-01T23:59:59.000Z

146

URANIUM IN ALKALINE ROCKS  

E-Print Network [OSTI]

Greenland," in Uranium Exploration Geology, Int. AtomicOklahoma," 1977 Nure Geology Uranium Symposium, Igneous HostMcNeil, M. , 1977. "Geology of Brazil's Uranium and Thorium

Murphy, M.

2011-01-01T23:59:59.000Z

147

Investigation of Trace Uranium in Biological Matrices  

E-Print Network [OSTI]

complex. As a result, the data varies in its breadth and quality due to the variety of sources.[41-44] Additional studies have been undertaken to understand the effects of using depleted uranium munitions in war and the accompanying exposures.[45...

Miller, James Christopher

2013-05-31T23:59:59.000Z

148

Gas shielding apparatus  

DOE Patents [OSTI]

An apparatus for preventing oxidation by uniformly distributing inert shielding gas over the weld area of workpieces such as pipes being welded together. The apparatus comprises a chamber and a gas introduction element. The chamber has an annular top wall, an annular bottom wall, an inner side wall and an outer side wall connecting the top and bottom walls. One side wall is a screen and the other has a portion defining an orifice. The gas introduction element has a portion which encloses the orifice and can be one or more pipes. The gas introduction element is in fluid communication with the chamber and introduces inert shielding gas into the chamber. The inert gas leaves the chamber through the screen side wall and is dispersed evenly over the weld area.

Brandt, D.

1984-06-05T23:59:59.000Z

149

Estimation of the Performance of Multiple Active Neutron Interrogation Signatures for Detecting Shielded HEU  

SciTech Connect (OSTI)

A comprehensive modeling study has been carried out to evaluate the utility of multiple active neutron interrogation signatures for detecting shielded highly enriched uranium (HEU). The modeling effort focused on varying HEU masses from 1 kg to 20 kg; varying types of shields including wood, steel, cement, polyethylene, and borated polyethylene; varying depths of the HEU in the shields, and varying engineered shields immediately surrounding the HEU including steel, tungsten, and cadmium. Neutron and gamma-ray signatures were the focus of the study and false negative detection probabilities versus measurement time were used as a performance metric. To facilitate comparisons among different approaches an automated method was developed to generate receiver operating characteristic (ROC) curves for different sets of model variables for multiple background count rate conditions. This paper summarizes results or the analysis, including laboratory benchmark comparisons between simulations and experiments. The important impact engineered shields can play towards degrading detectability and methods for mitigating this will be discussed.

David L. Chichester; Scott J. Thompson; Scott M. Watson; James T. Johnson; Edward H. Seabury

2012-10-01T23:59:59.000Z

150

The non-aqueous chemistry of uranium has been an active area of exploration in recent decades1,2  

E-Print Network [OSTI]

-purity depleted uranium produced as a by-product of nuclear isotope enrichment programmes. The early actinideThe non-aqueous chemistry of uranium has been an active area of exploration in recent decades1 for uranium will be created in part by the quest of researchers to understand the properties and potential

Cai, Long

151

Actively driven thermal radiation shield  

DOE Patents [OSTI]

A thermal radiation shield for cooled portable gamma-ray spectrometers. The thermal radiation shield is located intermediate the vacuum enclosure and detector enclosure, is actively driven, and is useful in reducing the heat load to mechanical cooler and additionally extends the lifetime of the mechanical cooler. The thermal shield is electrically-powered and is particularly useful for portable solid-state gamma-ray detectors or spectrometers that dramatically reduces the cooling power requirements. For example, the operating shield at 260K (40K below room temperature) will decrease the thermal radiation load to the detector by 50%, which makes possible portable battery operation for a mechanically cooled Ge spectrometer.

Madden, Norman W. (Livermore, CA); Cork, Christopher P. (Pleasant Hill, CA); Becker, John A. (Alameda, CA); Knapp, David A. (Livermore, CA)

2002-01-01T23:59:59.000Z

152

Session 9.3: Advances in Depleted Uranium Technology  

E-Print Network [OSTI]

The submitted manuscript has been authored by a contractor of the U.S. Government under contract DE-AC05-00OR22725. Accordingly, the U.S. Government retains a nonexclusive, royalty-free license to publish or reproduce the published form of this contribution, or allow others to do so, for U.S. Government purposes.

Robert R. Price; M. Jonathan Haire; Allen G. Croff; Robert R. Price; M. Jonathan Haire; Allen G. Croff

2001-01-01T23:59:59.000Z

153

DOE Issues Request for Quotations for Depleted Uranium Hexafluoride  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana. DOCUMENTSof Energy DOE ChallengeThese(Notice

154

Depleted Uranium Hexafluoride (DUF6) Fully Operational at the Portsmouth  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector General Office of Audit Services Audit ReportNextConditionalDepartment Federaland Paducah Gaseous

155

Background Fact Sheet Transfer of Depleted Uranium and Subsequent Transactions  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustionImprovement3--Logistical5/08 Attendance List1-02EvaluationJohn D.Department

156

LBNL Totally Depleted CCD Group Internal Note Last revised 8 November 2002  

E-Print Network [OSTI]

LBNL Totally Depleted CCD Group Internal Note Last revised 8 November 2002 Conceptual design for shielding the 200 µm-thick LBNL CCD at the Lick 3-m Coud´e Spectrograph from environmental gamma radiation This note was drafted by Don Groom (LBNL; deg@lbl.gov), but it is based on considerable work by Steve

157

LBNL Totally Depleted CCD Group Internal Note Last revised 8 November 2002  

E-Print Network [OSTI]

LBNL Totally Depleted CCD Group Internal Note Last revised 8 November 2002 Conceptual design for shielding the 200 µm­thick LBNL CCD at the Lick 3­m Coudâ??e Spectrograph from environmental gamma radiation This note was drafted by Don Groom (LBNL; deg@lbl.gov), but it is based on considerable work by Steve

158

Drip Shield Emplacement Gantry Concept  

SciTech Connect (OSTI)

This design analysis has shown that, on a conceptual level, the emplacement of drip shields is feasible with current technology and equipment. A plan for drip shield emplacement was presented using a Drip Shield Transporter, a Drip Shield Emplacement Gantry, a locomotive, and a Drip Shield Gantry Carrier. The use of a Drip Shield Emplacement Gantry as an emplacement concept results in a system that is simple, reliable, and interfaces with the numerous other exising repository systems. Using the Waste Emplacement/Retrieval System design as a basis for the drip shield emplacement concept proved to simplify the system by using existing equipment, such as the gantry carrier, locomotive, Electrical and Control systems, and many other systems, structures, and components. Restricted working envelopes for the Drip Shield Emplacement System require further consideration and must be addressed to show that the emplacement operations can be performed as the repository design evolves. Section 6.1 describes how the Drip Shield Emplacement System may use existing equipment. Depending on the length of time between the conclusion of waste emplacement and the commencement of drip shield emplacement, this equipment could include the locomotives, the gantry carrier, and the electrical, control, and rail systems. If the exisiting equipment is selected for use in the Drip Shield Emplacement System, then the length of time after the final stages of waste emplacement and start of drip shield emplacement may pose a concern for the life cycle of the system (e.g., reliability, maintainability, availability, etc.). Further investigation should be performed to consider the use of existing equipment for drip shield emplacement operations. Further investigation will also be needed regarding the interfaces and heat transfer and thermal effects aspects. The conceptual design also requires further design development. Although the findings of this analysis are accurate for the assumptions made, further refinements of this analysis are needed as the project parameters change. The designs of the drip shield, the Emplacement Drift, and the other drip shield emplacement equipment all have a direct effect on the overall design feasibility.

Silva, R.A.; Cron, J.

2000-03-29T23:59:59.000Z

159

The uranium cylinder assay system for enrichment plant safeguards  

SciTech Connect (OSTI)

Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component in preventing the spread of nuclear weapons. A useful tool for the nuclear materials accountancy of such a plant would be an instrument that measured the uranium content of UF{sub 6} cylinders. The Uranium Cylinder Assay System (UCAS) was designed for Japan Nuclear Fuel Limited (JNFL) for use in the Rokkasho Enrichment Plant in Japan for this purpose. It uses total neutron counting to determine uranium mass in UF{sub 6} cylinders given a known enrichment. This paper describes the design of UCAS, which includes features to allow for unattended operation. It can be used on 30B and 48Y cylinders to measure depleted, natural, and enriched uranium. It can also be used to assess the amount of uranium in decommissioned equipment and waste containers. Experimental measurements have been carried out in the laboratory and these are in good agreement with the Monte Carlo modeling results.

Miller, Karen A [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Marlow, Johnna B [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Rael, Carlos D [Los Alamos National Laboratory; Iwamoto, Tomonori [JNFL; Tamura, Takayuki [JNFL; Aiuchi, Syun [JNFL

2010-01-01T23:59:59.000Z

160

a. ASTM Standard C787-11, Standard Specification for Uranium...  

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

in support of a request for proposals to design, build, and operate facilities to convert depleted uranium hexafluoride (DUF 6 ) to more chemically stable forms. On page C-8 in the...

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Uranium Certified Reference Materials Price List | U.S. DOE Office...  

Office of Science (SC) Website

Hexafluoride (4.5% U-235) 1700 g 59,420 . .pdf file (50KB) . .pdf file (63KB) A 115 Uranium (Depleted) Metal (0.99977 g Ug) 75 g 2,980 . .pdf file (121KB) . .pdf file...

162

Portable convertible blast effects shield  

DOE Patents [OSTI]

A rapidly deployable portable convertible blast effects shield/ballistic shield includes a set two or more frusto-conically-tapered telescoping rings operably connected to each other to convert between a telescopically-collapsed configuration for storage and transport, and a telescopically-extended upright configuration forming an expanded inner volume. In a first embodiment, the upright configuration provides blast effects shielding, such as against blast pressures, shrapnel, and/or fire balls. And in a second embodiment, the upright configuration provides ballistic shielding, such as against incoming weapons fire, shrapnel, etc. Each ring has a high-strength material construction, such as a composite fiber and matrix material, capable of substantially inhibiting blast effects and impinging projectiles from passing through the shield. And the set of rings are releasably securable to each other in the telescopically-extended upright configuration by the friction fit of adjacent pairs of frusto-conically-tapered rings to each other.

Pastrnak, John W. (Livermore, CA); Hollaway, Rocky (Modesto, CA); Henning, Carl D. (Livermore, CA); Deteresa, Steve (Livermore, CA); Grundler, Walter (Hayward, CA); Hagler, Lisle B. (Berkeley, CA); Kokko, Edwin (Dublin, CA); Switzer, Vernon A. (Livermore, CA)

2011-03-15T23:59:59.000Z

163

Portable convertible blast effects shield  

DOE Patents [OSTI]

A rapidly deployable portable convertible blast effects shield/ballistic shield includes a set two or more telescoping cylindrical rings operably connected to each other to convert between a telescopically-collapsed configuration for storage and transport, and a telescopically-extended upright configuration forming an expanded inner volume. In a first embodiment, the upright configuration provides blast effects shielding, such as against blast pressures, shrapnel, and/or fire balls. And in a second embodiment, the upright configuration provides ballistic shielding, such as against incoming weapons fire, shrapnel, etc. Each ring has a high-strength material construction, such as a composite fiber and matrix material, capable of substantially inhibiting blast effects and impinging projectiles from passing through the shield. And the set of rings are releasably securable to each other in the telescopically-extended upright configuration, such as by click locks.

Pastrnak, John W. (Livermore, CA); Hollaway, Rocky (Modesto, CA); Henning, Carl D. (Livermore, CA); Deteresa, Steve (Livermore, CA); Grundler, Walter (Hayward, CA); Hagler, Lisle B. (Berkeley, CA); Kokko, Edwin (Dublin, CA); Switzer, Vernon A (Livermore, CA)

2007-05-22T23:59:59.000Z

164

Portable convertible blast effects shield  

DOE Patents [OSTI]

A rapidly deployable portable convertible blast effects shield/ballistic shield includes a set two or more telescoping cylindrical rings operably connected to each other to convert between a telescopically-collapsed configuration for storage and transport, and a telescopically-extended upright configuration forming an expanded inner volume. In a first embodiment, the upright configuration provides blast effects shielding, such as against blast pressures, shrapnel, and/or fire balls. And in a second embodiment, the upright configuration provides ballistic shielding, such as against incoming weapons fire, shrapnel, etc. Each ring has a high-strength material construction, such as a composite fiber and matrix material, capable of substantially inhibiting blast effects and impinging projectiles from passing through the shield. And the set of rings are releasably securable to each other in the telescopically-extended upright configuration, such as by click locks.

Pastrnak, John W. (Livermore, CA); Hollaway, Rocky (Modesto, CA); Henning, Carl D. (Livermore, CA); Deteresa, Steve (Livermore, CA); Grundler, Walter (Hayward, CA); Hagler,; Lisle B. (Berkeley, CA); Kokko, Edwin (Dublin, CA); Switzer, Vernon A (Livermore, CA)

2010-10-26T23:59:59.000Z

165

Welding shield for coupling heaters  

DOE Patents [OSTI]

Systems for coupling end portions of two elongated heater portions and methods of using such systems to treat a subsurface formation are described herein. A system may include a holding system configured to hold end portions of the two elongated heater portions so that the end portions are abutted together or located near each other; a shield for enclosing the end portions, and one or more inert gas inlets configured to provide at least one inert gas to flush the system with inert gas during welding of the end portions. The shield may be configured to inhibit oxidation during welding that joins the end portions together. The shield may include a hinged door that, when closed, is configured to at least partially isolate the interior of the shield from the atmosphere. The hinged door, when open, is configured to allow access to the interior of the shield.

Menotti, James Louis (Dickinson, TX)

2010-03-09T23:59:59.000Z

166

Uranium industry annual 1996  

SciTech Connect (OSTI)

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.

NONE

1997-04-01T23:59:59.000Z

167

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

Broader source: Energy.gov [DOE]

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 materials to a disposal facility; transportation and sale of the hydrogen fluoride (HF) produced as a conversion co-product; and neutralization of HF to calcium fluoride and its sale or disposal in the event that the HF product is not sold.

168

Bugs boost Cold War clean-up: Bacteria could scrub uranium from sites contaminated decades ago. updated at midnight GMTtoday is friday, november 14  

E-Print Network [OSTI]

2003 · Fungus catches radioactive fallout 8 May 2002 · Depleted uranium soils battlefields 12 MarchBugs boost Cold War clean-up: Bacteria could scrub uranium from sites contaminated decades ago boost Cold War clean-up Bacteria could scrub uranium from sites contaminated decades ago. 13 October

Lovley, Derek

169

Radiation Shielding and Radiological Protection  

E-Print Network [OSTI]

Radiation Shielding and Radiological Protection J. Kenneth Shultis Richard E. Faw Department@triad.rr.com Radiation Fields and Sources ................................................ . Radiation Field Variables........................................................... .. Direction and Solid Angle Conventions ......................................... .. Radiation Fluence

Shultis, J. Kenneth

170

Thermal neutron shield and method of manufacture  

DOE Patents [OSTI]

A thermal neutron shield comprising boron shielding panels with a high percentage of the element Boron. The panel is least 46% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of boron shielding panels which includes enriching the pre-cursor mixture with varying grit sizes of Boron Carbide.

Metzger, Bert Clayton; Brindza, Paul Daniel

2014-03-04T23:59:59.000Z

171

Uranium Industry Annual, 1992  

SciTech Connect (OSTI)

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.

Not Available

1993-10-28T23:59:59.000Z

172

SCALE Continuous-Energy Monte Carlo Depletion with Parallel KENO in TRITON  

SciTech Connect (OSTI)

The TRITON sequence of the SCALE code system is a powerful and robust tool for performing multigroup (MG) reactor physics analysis using either the 2-D deterministic solver NEWT or the 3-D Monte Carlo transport code KENO. However, as with all MG codes, the accuracy of the results depends on the accuracy of the MG cross sections that are generated and/or used. While SCALE resonance self-shielding modules provide rigorous resonance self-shielding, they are based on 1-D models and therefore 2-D or 3-D effects such as heterogeneity of the lattice structures may render final MG cross sections inaccurate. Another potential drawback to MG Monte Carlo depletion is the need to perform resonance self-shielding calculations at each depletion step for each fuel segment that is being depleted. The CPU time and memory required for self-shielding calculations can often eclipse the resources needed for the Monte Carlo transport. This summary presents the results of the new continuous-energy (CE) calculation mode in TRITON. With the new capability, accurate reactor physics analyses can be performed for all types of systems using the SCALE Monte Carlo code KENO as the CE transport solver. In addition, transport calculations can be performed in parallel mode on multiple processors.

Goluoglu, Sedat [ORNL] [ORNL; Bekar, Kursat B [ORNL] [ORNL; Wiarda, Dorothea [ORNL] [ORNL

2012-01-01T23:59:59.000Z

173

Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-zirconium Alloys for Advanced Nuclear Fuel Applications  

E-Print Network [OSTI]

The research in this thesis covers the design and implementation of a depleted uranium (DU) powder production system and the initial results of a DU-Zr-Mg alloy alpha phase sintering experiment where the Mg is a surrogate for Pu and Am. The powder...

Garnetti, David J.

2010-07-14T23:59:59.000Z

174

San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses  

SciTech Connect (OSTI)

The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

Hermann, O.W.

1999-09-01T23:59:59.000Z

175

Atlas SCT/Pixel Grounding and Shielding ATLAS SCT/Pixel Grounding and Shielding Note  

E-Print Network [OSTI]

Atlas SCT/Pixel Grounding and Shielding 1 ATLAS SCT/Pixel Grounding and Shielding Note November 22 mostly connects existing mechanical electrical conductive #12; Atlas SCT/Pixel Grounding and Shielding 2 that equivalent. The barrel outer heat shield (150 aluminum) main element shield. #12; Atlas SCT/Pixel Grounding

California at Santa Cruz, University of

176

Shielding vacuum fluctuations with graphene  

E-Print Network [OSTI]

The Casimir-Polder interaction of ground-state and excited atoms with graphene is investigated with the aim to establish whether graphene systems can be used as a shield for vacuum fluctuations of an underlying substrate. We calculate the zero-temperature Casimir-Polder potential from the reflection coefficients of graphene within the framework of the Dirac model. For both doped and undoped graphene we show limits at which graphene could be used effectively as a shield. Additional results are given for AB-stacked bilayer graphene.

Sofia Ribeiro; Stefan Scheel

2014-03-14T23:59:59.000Z

177

Uranium industry annual 1994  

SciTech Connect (OSTI)

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.

NONE

1995-07-05T23:59:59.000Z

178

Fully depleted back illuminated CCD  

DOE Patents [OSTI]

A backside illuminated charge coupled device (CCD) is formed of a relatively thick high resistivity photon sensitive silicon substrate, with frontside electronic circuitry, and an optically transparent backside ohmic contact for applying a backside voltage which is at least sufficient to substantially fully deplete the substrate. A greater bias voltage which overdepletes the substrate may also be applied. One way of applying the bias voltage to the substrate is by physically connecting the voltage source to the ohmic contact. An alternate way of applying the bias voltage to the substrate is to physically connect the voltage source to the frontside of the substrate, at a point outside the depletion region. Thus both frontside and backside contacts can be used for backside biasing to fully deplete the substrate. Also, high resistivity gaps around the CCD channels and electrically floating channel stop regions can be provided in the CCD array around the CCD channels. The CCD array forms an imaging sensor useful in astronomy.

Holland, Stephen Edward (Hercules, CA)

2001-01-01T23:59:59.000Z

179

Rotational Mixing and Lithium Depletion  

E-Print Network [OSTI]

I review basic observational features in Population I stars which strongly implicate rotation as a mixing agent; these include dispersion at fixed temperature in coeval populations and main sequence lithium depletion for a range of masses at a rate which decays with time. New developments related to the possible suppression of mixing at late ages, close binary mergers and their lithium signature, and an alternate origin for dispersion in young cool stars tied to radius anomalies observed in active young stars are discussed. I highlight uncertainties in models of Population II lithium depletion and dispersion related to the treatment of angular momentum loss. Finally, the origins of rotation are tied to conditions in the pre-main sequence, and there is thus some evidence that enviroment and planet formation could impact stellar rotational properties. This may be related to recent observational evidence for cluster to cluster variations in lithium depletion and a connection between the presence of planets and s...

Pinsonneault, M H

2010-01-01T23:59:59.000Z

180

Nuclear power fleets and uranium resources recovered from phosphates  

SciTech Connect (OSTI)

Current light water reactors (LWR) burn fissile uranium, whereas some future reactors, as Sodium fast reactors (SFR) will be capable of recycling their own plutonium and already-extracted depleted uranium. This makes them a feasible solution for the sustainable development of nuclear energy. Nonetheless, a sufficient quantity of plutonium is needed to start up an SFR, with the plutonium already being produced in light water reactors. The availability of natural uranium therefore has a direct impact on the capacity of the reactors (both LWR and SFR) that we can build. It is therefore important to have an accurate estimate of the available uranium resources in order to plan for the world's future nuclear reactor fleet. This paper discusses the correspondence between the resources (uranium and plutonium) and the nuclear power demand. Sodium fast reactors will be built in line with the availability of plutonium, including fast breeders when necessary. Different assumptions on the global uranium resources are taken into consideration. The largely quoted estimate of 22 Mt of uranium recovered for phosphate rocks can be seriously downscaled. Based on our current knowledge of phosphate resources, 4 Mt of recoverable uranium already seems to be an upper bound value. The impact of the downscaled estimate on the deployment of a nuclear fleet is assessed accordingly. (authors)

Gabriel, S.; Baschwitz, A.; Mathonniere, G. [CEA, DEN/DANS/I-tese, F-91191 Gif-sur-Yvette (France)

2013-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Penetration seals for TFTR shielding  

SciTech Connect (OSTI)

The penetrations of the shielding provided for TFTR are required to be sealed to avoid radiation streaming. This report provides a discussion of the properties required for these penetration seals. Several alternate designs are discussed and evaluated and designs recommended for specific applications.

Hondorp, H.L.

1980-12-01T23:59:59.000Z

182

Uranium in the Oatman Creek granite of Central Texas and its economic potential  

E-Print Network [OSTI]

, however, the need to explore for new materials containing uranium will incr ease as the high grade sedimentary uranium deposits become depleted. A logical place to begin this search lies with the source rock for many of the known sedimentary uranium... potential uranium source. Th1s study focuses on an 80 acre outcrop of the Oatman Creek granite known as Bear Mountain, in Gillespie County, Texas. The gran1te is a medium-grained, gray to pink rock. Nodal analysis indicates the composit1on 1s 35. 5...

Conrad, Curtis Paul

1982-01-01T23:59:59.000Z

183

Shielding and grounding in large detectors  

SciTech Connect (OSTI)

Prevention of electromagnetic interference (EMI), or ``noise pickup,`` is an important design aspect in large detectors in accelerator environments. Shielding effectiveness as a function of shield thickness and conductivity vs the type and frequency of the interference field is described. Noise induced in transmission lines by ground loop driven currents in the shield is evaluated and the importance of low shield resistance is emphasized. Some measures for prevention of ground loops and isolation of detector-readout systems are discussed.

Radeka, V.

1998-09-01T23:59:59.000Z

184

Experimental partitioning of uranium between liquid iron sulfide and liquid silicate: Implications for radioactivity in the Earth's core  

E-Print Network [OSTI]

Experimental partitioning of uranium between liquid iron sulfide and liquid silicate: Implications Measurable uranium (U) is found in metal sulfide liquids in equilibrium with molten silicate at conditions shows that K is depleted in the Earth by $50%, while U and Th are slightly enriched (Palme and O

Minarik, William

185

Biological Diversity of the Guiana Shield: Georeferencing Plants of the Guiana Shield  

E-Print Network [OSTI]

of Suriname, maximum elevation 500 m] #12;Georeferencing Plants of the Guiana Shield Google Earth allows

Mathis, Wayne N.

186

Modified shielding jet model for twin-jet shielding analysis  

E-Print Network [OSTI]

the slowing of the jet flow due to turbulent mixing and entrainment of particles from the surrounding medium. The empirical formulations and velocity profiles derived for the respective regions of the jet consider this increase in entrained fluid... velocity profiles are integrated over their respective cross sections of the shielding jet to determine the total volumetric flowrate at the specified locations. A slug flow velocity approximation is then determined for each of the desired downstream...

Gilbride, Jennifer Frances

2012-06-07T23:59:59.000Z

187

Final Uranium Leasing Program Programmatic Environmental Impact...  

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

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

188

IPNS enriched uranium booster target  

SciTech Connect (OSTI)

Since startup in 1981, IPNS has operated on a fully depleted /sup 238/U target. With the booster as in the present system, high energy protons accelerated to 450 MeV by the Rapid Cycling Synchrotron are directed at the target and by mechanisms of spallation and fission of the uranium, produce fast neutrons. The neutrons from the target pass into adjacent moderator where they slow down to energies useful for spectroscopy. The target cooling systems and monitoring systems have operated very reliably and safely during this period. To provide higher neutron intensity, we have developed plans for an enriched uranium (booster) target. HETC-VIM calculations indicate that the target will produce approx.90 kW of heat, with a nominal x5 gain (k/sub eff/ = 0.80). The neutron beam intensity gain will be a factor of approx.3. Thermal-hydraulic and heat transport calculations indicate that approx.1/2 in. thick /sup 235/U discs are subject to about the same temperatures as the present /sup 238/U 1 in. thick discs. The coolant will be light demineralized water (H/sub 2/O) and the coolant flow rate must be doubled. The broadening of the fast neutron pulse width should not seriously affect the neutron scattering experiments. Delayed neutrons will appear at a level about 3% of the total (currently approx.0.5%). This may affect backgrounds in some experiments, so that we are assessing measures to control and correct for this (e.g., beam tube choppers). Safety analyses and neutronic calculations are nearing completion. Construction of the /sup 235/U discs at the ORNL Y-12 facility is scheduled to begin late 1985. The completion of the booster target and operation are scheduled for late 1986. No enriched uranium target assembly operating at the projected power level now exists in the world. This effort thus represents an important technological experiment as well as being a ''flux enhancer''.

Schulke, A.W. Jr.

1985-01-01T23:59:59.000Z

189

Measurement and Analysis of Fission Rates in a Spherical Mockup of Uranium and Polyethylene  

E-Print Network [OSTI]

Measurements of the reaction rate distribution were carried out using two kinds of Plate Micro Fission Chamber(PMFC). The first is a depleted uranium chamber and the second an enriched uranium chamber. The material in the depleted uranium chamber is strictly the same as the material in the uranium assembly. With the equation solution to conduct the isotope contribution correction, the fission rate of 238U and 235U were obtained from the fission rate of depleted uranium and enriched uranium. And then, the fission count of 238U and 235U in an individual uranium shell was obtained. In this work, MCNP5 and continuous energy cross sections ENDF/BV.0 were used for the analysis of fission rate distribution and fission count. The calculated results were compared with the experimental ones. The calculation of fission rate of DU and EU were found to agree with the measured ones within 10% except at the positions in polyethylene region and the two positions near the outer surface. Beacause the fission chamber was not co...

Tong-Hua, Zhu; Xin-Xin, Lu; Rong, Liu; Zi-Jie, Han; Li, Jiang; Mei, Wang

2013-01-01T23:59:59.000Z

190

Light shield for solar concentrators  

DOE Patents [OSTI]

A solar receiver unit including a housing defining a recess, a cell assembly received in the recess, the cell assembly including a solar cell, and a light shield received in the recess and including a body and at least two tabs, the body defining a window therein, the tabs extending outward from the body and being engaged with the recess, wherein the window is aligned with the solar cell.

Plesniak, Adam P.; Martins, Guy L.

2014-08-26T23:59:59.000Z

191

Method for converting uranium oxides to uranium metal  

DOE Patents [OSTI]

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.

Duerksen, Walter K. (Norris, TN)

1988-01-01T23:59:59.000Z

192

Biological shield design and analysis of KIPT accelerator-driven subcritical facility.  

SciTech Connect (OSTI)

Argonne National Laboratory of the United States and Kharkov Institute of Physics and Technology of Ukraine have been collaborating on the conceptual design development of an electron accelerator-driven subcritical facility. The facility will be utilized for performing basic and applied nuclear research, producing medical isotopes, and training young nuclear specialists. This paper presents the design and analyses of the biological shield performed for the top section of the facility. The neutron source driving the subcritical assembly is generated from the interaction of a 100-kW electron beam with a natural uranium target. The electron energy is in the range of 100 to 200 MeV, and it has a uniform spatial distribution. The shield design and the associated analyses are presented including different parametric studies. In the analyses, a significant effort was dedicated to the accurate prediction of the radiation dose outside the shield boundary as a function of the shield thickness without geometrical approximations or material homogenization. The MCNPX Monte Carlo code was utilized for the transport calculation of electrons, photons, and neutrons. Weight window variance-reduction techniques were introduced, and the dose equivalent outside the shield can be calculated with reasonably good statistics.

Zhong, Z.; Gohar, Y.; Nuclear Engineering Division

2009-12-01T23:59:59.000Z

193

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

DOE Patents [OSTI]

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

Willit, James L. (Ratavia, IL)

2007-09-11T23:59:59.000Z

194

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

DOE Patents [OSTI]

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

Willit, James L. (Batavia, IL)

2010-09-21T23:59:59.000Z

195

Optical Constants ofOptical Constants of Uranium Nitride Thin FilmsUranium Nitride Thin Films  

E-Print Network [OSTI]

Optical Constants ofOptical Constants of Uranium Nitride Thin FilmsUranium Nitride Thin FilmsDelta--Beta Scatter Plot at 220 eVBeta Scatter Plot at 220 eV #12;Why Uranium Nitride?Why Uranium Nitride? UraniumUranium, uranium,Bombard target, uranium, with argon ionswith argon ions Uranium atoms leaveUranium atoms leave

Hart, Gus

196

Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.  

SciTech Connect (OSTI)

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division); (2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department)

2012-04-04T23:59:59.000Z

197

EPA Update: NESHAP Uranium Activities  

E-Print Network [OSTI]

for underground uranium mining operations (Subpart B) EPA regulatory requirements for operating uranium mill for Underground Uranium Mining Operations (Subpart B) #12;5 EPA Regulatory Requirements for Underground Uranium uranium mines include: · Applies to 10,000 tons/yr ore production, or 100,000 tons/mine lifetime · Ambient

198

Uranium hexafluoride public risk  

SciTech Connect (OSTI)

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

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

1994-08-01T23:59:59.000Z

199

Shield Volcano | Open Energy Information  

Open Energy Info (EERE)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to:Seadov Pty Ltd Jump to: navigation,Pvt Ltd Jump to:Shenzhen79. It.Shida BatteryShield

200

Uranium Mill Tailings Management  

SciTech Connect (OSTI)

This book presents the papers given at the Fifth Symposium on Uranium Mill Tailings Management. Advances made with regard to uranium mill tailings management, environmental effects, regulations, and reclamation are reviewed. Topics considered include tailings management and design (e.g., the Uranium Mill Tailings Remedial Action Project, environmental standards for uranium mill tailings disposal), surface stabilization (e.g., the long-term stability of tailings, long-term rock durability), radiological aspects (e.g. the radioactive composition of airborne particulates), contaminant migration (e.g., chemical transport beneath a uranium mill tailings pile, the interaction of acidic leachate with soils), radon control and covers (e.g., radon emanation characteristics, designing surface covers for inactive uranium mill tailings), and seepage and liners (e.g., hydrologic observations, liner requirements).

Nelson, J.D.

1982-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Investigation of breached depleted UF{sub 6} cylinders  

SciTech Connect (OSTI)

In June 1990, during a three-site inspection of cylinders being used for long-term storage of solid depleted UF{sub 6}, two 14-ton steel cylinders at Portsmouth, Ohio, were discovered with holes in the barrel section of the cylinders. Both holes, concealed by UF{sub 4} reaction products identical in color to the cylinder coating, were similarly located near the front stiffening ring. The UF{sub 4} appeared to have self-sealed the holes, thus containing nearly all of the uranium contents. Martin Marietta Energy Systems, Inc., Vice President K.W. Sommerfeld immediately formed an investigation team to: (1) identify the most likely cause of failure for the two breached cylinders, (2) determine the impact of these incidents on the three-site inventory, and (3) provide recommendations and preventive measures. This document discusses the results of this investigation.

Barber, E.J.; Butler, T.R.; DeVan, J.H.; Googin, J.M.; Taylor, M.S.; Dyer, R.H.; Russell, J.R.

1991-09-01T23:59:59.000Z

202

Investigation of breached depleted UF sub 6 cylinders  

SciTech Connect (OSTI)

In June 1990, during a three-site inspection of cylinders being used for long-term storage of solid depleted UF{sub 6}, two 14-ton steel cylinders at Portsmouth, Ohio, were discovered with holes in the barrel section of the cylinders. Both holes, concealed by UF{sub 4} reaction products identical in color to the cylinder coating, were similarly located near the front stiffening ring. The UF{sub 4} appeared to have self-sealed the holes, thus containing nearly all of the uranium contents. Martin Marietta Energy Systems, Inc., Vice President K.W. Sommerfeld immediately formed an investigation team to: (1) identify the most likely cause of failure for the two breached cylinders, (2) determine the impact of these incidents on the three-site inventory, and (3) provide recommendations and preventive measures. This document discusses the results of this investigation.

Barber, E.J.; Butler, T.R.; DeVan, J.H.; Googin, J.M.; Taylor, M.S.; Dyer, R.H.; Russell, J.R.

1991-09-01T23:59:59.000Z

203

Process for continuous production of metallic uranium and uranium alloys  

DOE Patents [OSTI]

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.

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

1995-06-06T23:59:59.000Z

204

Process for continuous production of metallic uranium and uranium alloys  

DOE Patents [OSTI]

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.

Hayden, Jr., Howard W. (Oakridge, TN); Horton, James A. (Livermore, CA); Elliott, Guy R. B. (Los Alamos, NM)

1995-01-01T23:59:59.000Z

205

Preparation of uranium compounds  

DOE Patents [OSTI]

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.

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

2013-02-19T23:59:59.000Z

206

Shielding effectiveness against electromagnetic interference  

SciTech Connect (OSTI)

The use of metal-filled and metal-coated plastics and other modified dielectric materials to replace metals for enclosures has created a need to test these materials for their electromagnetic interference (EMI) shielding effectiveness (SE). Shielding effectiveness involves a variety of electromagnetic environments, and useful data can be obtained from tests that carefully limit the environment to that of a plane wave. Such an environment can be created in a circular or rectangular transmission line. Two such transmission line test fixtures, which hold samples of the material to be tested, have been developed. The fixtures described in this report are the National Bureau of Standards (NBS) coaxial transverse electromagnetic (TEM) cell, and a dual TEM cell constructed at ORNL from a design suggested by the NBS. The NBS coaxial fixture is an improved version of the device recommended by the American Society for Testing and Materials (ASTM). The problems associated with measuring SE are well described in the literature. The two methods described here are the result of years of work to establish procedures and instrumentation that will produce acceptable data.

Googe, J.M.; Hess, R.A.

1987-10-01T23:59:59.000Z

207

Decontaminating lead bricks and shielding  

SciTech Connect (OSTI)

Lead used for shielding is often surface contaminated with radioisotopes and is therefore a RCRA D008 mixed waste. The technology-based standard for treatment is macroencapsulation. However, decontaminating and recycling the clean lead is a more attractive solution. Los Alamos National Laboratory decontaminates material and equipment contaminated with radioisotopes. Decontaminating lead poses special problems because of the RCRA hazard classification and the size of the inventory, now about 50 tons and likely to grow substantially of planned decommissioning operations. Thus lead, in the form of bricks and other shield shapes, is surface contaminated with fission products. One of the best methods for contaminated lead is removing the superficial layer of contamination with an abrasive medium under pressure. For lead, a mixture of alumina with water and air at about 40 psig rapidly and effectively decontaminates the lead. The abrasive medium is sprayed onto the lead in a scaled-off area. The slurry of abrasive and particles of lead falls through a floor and is collected in a sump. A pump sends the slurry mixture back to the spray gun, creating a continuous process. The process generates small volumes of lead slurry that can be solidified and, because it passes the TCLP, is not a mixed waste. The decontaminated lead can be released for recycling.

Lussiez, G.

1994-02-01T23:59:59.000Z

208

Decontaminating lead bricks and shielding  

SciTech Connect (OSTI)

Lead used for shielding is often surface contaminated with radioisotopes and is therefore a RCRA D008 mixed waste. The technology-based standard for treatment is macroencapsulation. However, decontaminating and recycling the clean lead is a more attractive solution. Los Alamos National Laboratory decontaminates material and equipment contaminated with radioisotopes. Decontaminating lead poses special problems because of the RCRA hazard classification and the size of the inventory, now about 50 tons and likely to grow substantially because of planned decommissioning operations. This lead, in the form of bricks and other shield shapes, is surface contaminated with fission products. One of the best methods for decontaminating lead is removing the thin superficial layer of contamination with an abrasive medium trader pressure. For lead, a mixture of alumina with water and air at about 40 psig rapidly and effectively decontaminates the lead. The abrasive medium is sprayed onto the lead in a sealed-off area. The slurry of abrasive and particles of lead falls through a floor grating and is collected in a sump. A pump sends the slurry mixture back to the spray gun, creating a continuous process. The process generates small volumes of contaminated lead slurry that can be solidified and, because it passes the TCLP, is not a mixed waste. The decontaminated lead can be released for recycling.

Lussiez, G.W.

1993-05-01T23:59:59.000Z

209

Decontaminating lead bricks and shielding  

SciTech Connect (OSTI)

Lead used for shielding is often surface contaminated with radioisotopes and is therefore a RCRA D008 mixed waste. The technology-based standard for treatment is macroencapsulation. However, decontaminating and recycling the clean lead is a more attractive solution. Los Alamos National Laboratory decontaminates material and equipment contaminated with radioisotopes. Decontaminating lead poses special problems because of the RCRA hazard classification and the size of the inventory, now about 50 tons and likely to grow substantially because of planned decommissioning operations. This lead, in the form of bricks and other shield shapes, is surface contaminated with fission products. One of the best methods for decontaminating lead is removing the thin superficial layer of contamination with an abrasive medium trader pressure. For lead, a mixture of alumina with water and air at about 40 psig rapidly and effectively decontaminates the lead. The abrasive medium is sprayed onto the lead in a sealed-off area. The slurry of abrasive and particles of lead falls through a floor grating and is collected in a sump. A pump sends the slurry mixture back to the spray gun, creating a continuous process. The process generates small volumes of contaminated lead slurry that can be solidified and, because it passes the TCLP, is not a mixed waste. The decontaminated lead can be released for recycling.

Lussiez, G.W.

1993-01-01T23:59:59.000Z

210

Lawrence Berkeley National Laboratory 1996 Site Environmental Report Vol. I  

E-Print Network [OSTI]

radioactive. uranium, depleted Uranium consisting primarilyoccurring in nature, depleted uranium is man-made. uranium,

2010-01-01T23:59:59.000Z

211

THE ENERGY SPECTRA OF URANIUM ATOMS SPUTTERED FROM URANIUM METAL AND URANIUM DIOXIDE TARGETS  

E-Print Network [OSTI]

THE ENERGY SPECTRA OF URANIUM ATOMS SPUTTERED FROM URANIUM METAL AND URANIUM DIOXIDE TARGETS Thesis. I have benefitted from conversations with many persons w~ile engaged in this project. I would like

Winfree, Erik

212

Uranium industry annual 1993  

SciTech Connect (OSTI)

Uranium production in the United States has declined dramatically from a peak of 43.7 million pounds U{sub 3}O{sub 8} (16.8 thousand metric tons uranium (U)) in 1980 to 3.1 million pounds U{sub 3}O{sub 8} (1.2 thousand metric tons U) in 1993. This decline is attributed to the world uranium market experiencing oversupply and intense competition. Large inventories of uranium accumulated when optimistic forecasts for growth in nuclear power generation were not realized. The other factor which is affecting U.S. uranium production is that some other countries, notably Australia and Canada, possess higher quality uranium reserves that can be mined at lower costs than those of the United States. Realizing its competitive advantage, Canada was the world`s largest producer in 1993 with an output of 23.9 million pounds U{sub 3}O{sub 8} (9.2 thousand metric tons U). The U.S. uranium industry, responding to over a decade of declining market prices, has downsized and adopted less costly and more efficient production methods. The main result has been a suspension of production from conventional mines and mills. Since mid-1992, only nonconventional production facilities, chiefly in situ leach (ISL) mining and byproduct recovery, have operated in the United States. In contrast, nonconventional sources provided only 13 percent of the uranium produced in 1980. ISL mining has developed into the most cost efficient and environmentally acceptable method for producing uranium in the United States. The process, also known as solution mining, differs from conventional mining in that solutions are used to recover uranium from the ground without excavating the ore and generating associated solid waste. This article describes the current ISL Yang technology and its regulatory approval process, and provides an analysis of the factors favoring ISL mining over conventional methods in a declining uranium market.

Not Available

1994-09-01T23:59:59.000Z

213

Thermal neutron shield and method of manufacture  

DOE Patents [OSTI]

A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

Brindza, Paul Daniel; Metzger, Bert Clayton

2013-05-28T23:59:59.000Z

214

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

SciTech Connect (OSTI)

This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials into pure HEU oxide and (2) blend the pure HEU oxide with depleted and natural uranium oxide to produce an LWR grade 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. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU will be produced as a waste suitable for storage or disposal.

NONE

1995-07-05T23:59:59.000Z

215

CHARACTERIZATION OF URANIUM, URANIUM OXIDE AND SILICON MULTILAYER THIN FILMS  

E-Print Network [OSTI]

CHARACTERIZATION OF URANIUM, URANIUM OXIDE AND SILICON MULTILAYER THIN FILMS by David T. Oliphant. Woolley Dean, College of Physical and Mathematical Sciences #12;ABSTRACT CHARACTERIZATION OF URANIUM, URANIUM OXIDE AND SILICON MULTILAYER THIN FILMS David T. Oliphant Department of Physics and Astronomy

Hart, Gus

216

Uranium dioxide electrolysis  

DOE Patents [OSTI]

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.

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

2009-12-29T23:59:59.000Z

217

Impact of External Heat-shielding Techniques on Shell Surface...  

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

External Heat-shielding Techniques on Shell Surface Temperatures and Dynamic Shell Thermal Deformation of Diesel Engine Emission Control Systems Impact of External Heat-shielding...

218

Idaho National Engineering and Environmental Laboratory Site Report on the Production and Use of Recycled Uranium  

SciTech Connect (OSTI)

Recent allegations regarding radiation exposure to radionuclides present in recycled uranium sent to the gaseous diffusion plants prompted the Department of Energy to undertake a system-wide study of recycled uranium. Of particular interest, were the flowpaths from site to site operations and facilities in which exposure to plutonium, neptunium and technetium could occur, and to the workers that could receive a significant radiation dose from handling recycled uranium. The Idaho National Engineering and Environmental Laboratory site report is primarily concerned with two locations. Recycled uranium was produced at the Idaho Chemical Processing Plant where highly enriched uranium was recovered from spent fuel. The other facility is the Specific Manufacturing Facility (SMC) where recycled, depleted uranium is manufactured into shapes for use by their customer. The SMC is a manufacturing facility that uses depleted uranium metal as a raw material that is then rolled and cut into shapes. There are no chemical processes that might concentrate any of the radioactive contaminant species. Recyclable depleted uranium from the SMC facility is sent to a private metallurgical facility for recasting. Analyses on the recast billets indicate that there is no change in the concentrations of transuranics as a result of the recasting process. The Idaho Chemical Processing Plant was built to recover high-enriched uranium from spent nuclear fuel from test reactors. The facility processed diverse types of fuel which required uniquely different fuel dissolution processes. The dissolved fuel was passed through three cycles of solvent extraction which resulted in a concentrated uranyl nitrate product. For the first half of the operating period, the uranium was shipped as the concentrated solution. For the second half of the operating period the uranium solution was thermally converted to granular, uranium trioxide solids. The dose reconstruction project has evaluated work exposure and exposure to the public as the result of normal operations and accidents that occurred at the INEEL. As a result of these studies, the maximum effective dose equivalent from site activities did not exceed seventeen percent of the natural background in Eastern Idaho. There was no year in which the radiation dose to the public exceeded the applicable limits for that year. Worker exposure to recycled uranium was minimized by engineering features that reduced the possibility of direct exposure.

L. C. Lewis; D. C. Barg; C. L. Bendixsen; J. P. Henscheid; D. R. Wenzel; B. L. Denning

2000-09-01T23:59:59.000Z

219

India's Worsening Uranium Shortage  

SciTech Connect (OSTI)

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.

Curtis, Michael M.

2007-01-15T23:59:59.000Z

220

Study of Active Shielding for {gamma} - Spectrometers  

SciTech Connect (OSTI)

The features of the ground located gamma ray spectrometer shielded passively with 12 cm of lead and actively by five 0.5m x 0.5m x 0.05m plastic veto shields are described. The detector mass related background was 0.345 C/kg s. The 511 keV annihilation line was reduced by the factor of 7 by the anticoincidence gate. It is shown that the plastic shields increase the neutron capture gamma line intensities due to neutron thermalization.

Bikit, I.; Mrdja, D.; Forkapic, S.; Todorovic, N.; Veskovic, M.; Slivka, J.; Conkic, Lj.; Krmar, M.; Varga, E. [Department of Physics, Faculty of Sciences, University of Novi Sad, Trg Dositeja Obradovica 4, 21 000 Novi Sad (Serbia and Montenegro)

2006-04-26T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Hot cell shield plug extraction apparatus  

DOE Patents [OSTI]

An apparatus is provided for moving shielding plugs into and out of holes in concrete shielding walls in hot cells for handling radioactive materials without the use of external moving equipment. The apparatus provides a means whereby a shield plug is extracted from its hole and then swung approximately 90 degrees out of the way so that the hole may be accessed. The apparatus uses hinges to slide the plug in and out and to rotate it out of the way, the hinge apparatus also supporting the weight of the plug in all positions, with the load of the plug being transferred to a vertical wall by means of a bolting arrangement.

Knapp, Philip A. (Moore, ID); Manhart, Larry K. (Pingree, ID)

1995-01-01T23:59:59.000Z

222

Seismic Crystals And Earthquake Shield Application  

E-Print Network [OSTI]

We theoretically demonstrate that earthquake shield made of seismic crystal can damp down surface waves, which are the most destructive type for constructions. In the paper, seismic crystal is introduced in aspect of band gaps (Stop band) and some design concepts for earthquake and tsunami shielding were discussed in theoretical manner. We observed in our FDTD based 2D elastic wave simulations that proposed earthquake shield could provide about 0.5 reductions in magnitude of surface wave on the Richter scale. This reduction rate in magnitude can considerably reduce destructions in the case of earthquake.

B. Baykant Alagoz; Serkan Alagoz

2009-05-15T23:59:59.000Z

223

Electromagnetic interference shielding using continuous carbon-fiber carbon-matrix and polymer-matrix composites  

E-Print Network [OSTI]

Electromagnetic interference shielding using continuous carbon-fiber carbon-matrix and polymer electromagnetic interference (EMI) shielding material with shielding effectiveness 124 dB, low surface impedance interference shielding 1. Introduction Electromagnetic interference (EMI) shielding is receiv- ing increasing

Chung, Deborah D.L.

224

SHIELDED CONTAINER COMPLETENESS COMMENTS July 13, 2010  

E-Print Network [OSTI]

the possible dose rate changes at the surface of the package for the intended payloads to be shipped. The same concerns as outlined above apply here. Reference WTS 2008. Shielded Container Type A Evaluation Report, ECO

225

Nuclear reactor shield including magnesium oxide  

DOE Patents [OSTI]

An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

1981-01-01T23:59:59.000Z

226

Method for the recovery of uranium values from uranium tetrafluoride  

DOE Patents [OSTI]

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.

Kreuzmann, Alvin B. (Cincinnati, OH)

1983-01-01T23:59:59.000Z

227

Method for the recovery of uranium values from uranium tetrafluoride  

DOE Patents [OSTI]

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.

Kreuzmann, A.B.

1982-10-27T23:59:59.000Z

228

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

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNationalRestart of the Review of theOFFICEACME |Supplement5869 Federal

229

Atlas SCT/Pixel Grounding and Shielding Note 1 ATLAS SCT/Pixel Grounding and Shielding Note  

E-Print Network [OSTI]

Atlas SCT/Pixel Grounding and Shielding Note 1 ATLAS SCT/Pixel Grounding and Shielding Note for SCT. This proposal mostly connects existing mechanical and electrical conductive #12;Atlas SCT. The barrel outer heat shield (150 µm aluminum) is the main element of the shield. #12;Atlas SCT

California at Santa Cruz, University of

230

Advancements in FBR shielding - Ten years in Japan  

SciTech Connect (OSTI)

Research and development in the area of fast breeder reactor (FBR) shielding in Japan was fully under way in April 1987 when criticality of the JOYO experimental FBR was first attained. The main activities performed and results obtained during more than 10 yr of FBR shielding research are presented. The paper describes shielding research in Joyo; Monju shielding design and related research; research activities for future FBRs; and evaluation of Monju shielding designs.

Ohtani, Nobuo; Suzuki, Soju

1990-01-01T23:59:59.000Z

231

Space Shielding Materials for Prometheus Application  

SciTech Connect (OSTI)

At the time of Prometheus program restructuring, shield material and design screening efforts had progressed to the point where a down-selection from approximately eighty-eight materials to a set of five ''primary'' materials was in process. The primary materials were beryllium (Be), boron carbide (B{sub 4}C), tungsten (W), lithium hydride (LiH), and water (H{sub 2}O). The primary materials were judged to be sufficient to design a Prometheus shield--excluding structural and insulating materials, that had not been studied in detail. The foremost preconceptual shield concepts included: (1) a Be/B{sub 4}C/W/LiH shield; (2) a Be/B{sub 4}C/W shield; (3) and a Be/B{sub 4}C/H{sub 2}O shield. Since the shield design and materials studies were still preliminary, alternative materials (e.g., {sup nal}B or {sup 10}B metal) were still being screened, but at a low level of effort. Two competing low mass neutron shielding materials are included in the primary materials due to significant materials uncertainties in both. For LiH, irradiation-induced swelling was the key issue, whereas for H{sub 2}O, containment corrosion without active chemistry control was key, Although detailed design studies are required to accurately estimate the mass of shields based on either hydrogenous material, both are expected to be similar in mass, and lower mass than virtually any alternative. Unlike Be, W, and B{sub 4}C, which are not expected to have restrictive temperature limits, shield temperature limits and design accommodations are likely to be needed for either LiH or H{sub 2}O. The NRPCT focused efforts on understanding swelting of LiH, and observed, from approximately fifty prior irradiation tests, that either casting ar thorough out-gassing should reduce swelling. A potential contributor to LiH swelling appears to be LiOH contamination due to exposure to humid air, that can be eliminated by careful processing. To better understand LiH irradiation performance and mitigate the risks in LiH development for a project with an aggressive schedule like JIMO, some background or advanced development effort for LiH should be considered for future space reactor projects.

R. Lewis

2006-01-20T23:59:59.000Z

232

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

DOE Patents [OSTI]

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.

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

1995-05-02T23:59:59.000Z

233

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

DOE Patents [OSTI]

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.

McLean, II, William (Oakland, CA); Miller, Philip E. (Livermore, CA); Horton, James A. (Livermore, CA)

1995-01-01T23:59:59.000Z

234

Uranium Enrichment Decontamination and Decommissioning Fund's...  

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

Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2008 and 2007 Financial Statement Audit Uranium Enrichment Decontamination and Decommissioning Fund's...

235

Process for electrolytically preparing uranium metal  

DOE Patents [OSTI]

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.

Haas, Paul A. (Knoxville, TN)

1989-01-01T23:59:59.000Z

236

Influence of uranium hydride oxidation on uranium metal behaviour  

SciTech Connect (OSTI)

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)

Patel, N.; Hambley, D. [National Nuclear Laboratory (United Kingdom); Clarke, S.A. [Sellafield Ltd (United Kingdom); Simpson, K.

2013-07-01T23:59:59.000Z

237

SONY GXB5005 GPS RECEIVER DATA Without shield can Installed With shield can in place  

E-Print Network [OSTI]

SONY GXB5005 GPS RECEIVER DATA Without shield can Installed With shield can in place (for illustrative purposes only) The Sony GXB5005 GPS receiver is a miniature 12 channel GPS module with support for WAAS/EGNOS augmented positioning. The receiver is based on Sony's CXD2951 single-chip GPS receiver IC

Berns, Hans-Gerd

238

Uranium-titanium-niobium alloy  

DOE Patents [OSTI]

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.

Ludtka, Gail M. (Oak Ridge, TN); Ludtka, Gerard M. (Oak Ridge, TN)

1990-01-01T23:59:59.000Z

239

Uranium hexafluoride handling. Proceedings  

SciTech Connect (OSTI)

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.

Not Available

1991-12-31T23:59:59.000Z

240

Microscreen radiation shield for thermoelectric generator  

DOE Patents [OSTI]

The present invention provides a microscreen radiation shield which reduces radiative heat losses in thermoelectric generators such as sodium heat engines without reducing the efficiency of operation of such devices. The radiation shield is adapted to be interposed between a reaction zone and a means for condensing an alkali metal vapor in a thermoelectric generator for converting heat energy directly to electrical energy. The radiation shield acts to reflect infrared radiation emanating from the reaction zone back toward the reaction zone while permitting the passage of the alkali metal vapor to the condensing means. The radiation shield includes a woven wire mesh screen or a metal foil having a plurality of orifices formed therein. The orifices in the foil and the spacing between the wires in the mesh is such that radiant heat is reflected back toward the reaction zone in the interior of the generator, while the much smaller diameter alkali metal atoms such as sodium pass directly through the orifices or along the metal surfaces of the shield and through the orifices with little or no impedance.

Hunt, Thomas K. (Ann Arbor, MI); Novak, Robert F. (Farmington Hills, MI); McBride, James R. (Ypsilanti, MI)

1990-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Cosmic Ray Interactions in Shielding Materials  

SciTech Connect (OSTI)

This document provides a detailed study of materials used to shield against the hadronic particles from cosmic ray showers at Earth’s surface. This work was motivated by the need for a shield that minimizes activation of the enriched germanium during transport for the MAJORANA collaboration. The materials suitable for cosmic-ray shield design are materials such as lead and iron that will stop the primary protons, and materials like polyethylene, borated polyethylene, concrete and water that will stop the induced neutrons. The interaction of the different cosmic-ray components at ground level (protons, neutrons, muons) with their wide energy range (from kilo-electron volts to giga-electron volts) is a complex calculation. Monte Carlo calculations have proven to be a suitable tool for the simulation of nucleon transport, including hadron interactions and radioactive isotope production. The industry standard Monte Carlo simulation tool, Geant4, was used for this study. The result of this study is the assertion that activation at Earth’s surface is a result of the neutronic and protonic components of the cosmic-ray shower. The best material to shield against these cosmic-ray components is iron, which has the best combination of primary shielding and minimal secondary neutron production.

Aguayo Navarrete, Estanislao; Kouzes, Richard T.; Ankney, Austin S.; Orrell, John L.; Berguson, Timothy J.; Troy, Meredith D.

2011-09-08T23:59:59.000Z

242

EIA - Natural Gas Pipeline Network - Depleted Reservoir Storage...  

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

Depleted Reservoir Storage Configuration About U.S. Natural Gas Pipelines - Transporting Natural Gas based on data through 20072008 with selected updates Depleted Production...

243

Shielding analysis and design of the KIPT experimental neutron source facility of Ukraine.  

SciTech Connect (OSTI)

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility based on the use of an electron accelerator driven subcritical (ADS) facility [1]. The facility uses the existing electron accelerators of KIPT in Ukraine. The neutron source of the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and the electron energy in the range of 100 to 200 MeV, [2]. The main functions of the facility are the production of medical isotopes and the support of the Ukraine nuclear power industry. Reactor physics experiments and material performance characterization will also be carried out. The subcritical assembly is driven by neutrons generated by the electron beam interactions with the target material. A fraction of these neutrons has an energy above 50 MeV generated through the photo nuclear interactions. This neutron fraction is very small and it has an insignificant contribution to the subcritical assembly performance. However, these high energy neutrons are difficult to shield and they can be slowed down only through the inelastic scattering with heavy isotopes. Therefore the shielding design of this facility is more challenging relative to fission reactors. To attenuate these high energy neutrons, heavy metals (tungsten, iron, etc.) should be used. To reduce the construction cost, heavy concrete with 4.8 g/cm{sup 3} density is selected as a shielding material. The iron weight fraction in this concrete is about 0.6. The shape and thickness of the heavy concrete shield are defined to reduce the biological dose equivalent outside the shield to an acceptable level during operation. At the same time, special attention was give to reduce the total shield mass to reduce the construction cost. The shield design is configured to maintain the biological dose equivalent during operation {le} 0.5 mrem/h inside the subcritical hall, which is five times less than the allowable dose for working forty hours per week for 50 weeks per year. This study analyzed and designed the thickness and the shape of the radial and top shields of the neutron source based on the biological dose equivalent requirements inside the subcritical hall during operation. The Monte Carlo code MCNPX is selected because of its capabilities for transporting electrons, photons, and neutrons. Mesh based weight windows variance reduction technique is utilized to estimate the biological dose outside the shield with good statistics. A significant effort dedicated to the accurate prediction of the biological dose equivalent outside the shield boundary as a function of the shield thickness without geometrical approximations or material homogenization. The building wall was designed with ordinary concrete to reduce the biological dose equivalent to the public with a safety factor in the range of 5 to 20.

Zhong, Z.; Gohar, M. Y. A.; Naberezhnev, D.; Duo, J.; Nuclear Engineering Division

2008-10-31T23:59:59.000Z

244

Fabrication of Cerium Oxide and Uranium Oxide Microspheres for Space Nuclear Power Applications  

SciTech Connect (OSTI)

Cerium oxide and uranium oxide microspheres are being produced via an internal gelation sol-gel method to investigate alternative fabrication routes for space nuclear fuels. Depleted uranium and non-radioactive cerium are being utilized as surrogates for plutonium-238 (Pu-238) used in radioisotope thermoelectric generators and for enriched uranium required by nuclear thermal rockets. While current methods used to produce Pu-238 fuels at Los Alamos National Laboratory (LANL) involve the generation of fine powders that pose a respiratory hazard and have a propensity to contaminate glove boxes, the sol-gel route allows for the generation of oxide microsphere fuels through an aqueous route. The sol-gel method does not generate fine powders and may require fewer processing steps than the LANL method with less operator handling. High-quality cerium dioxide microspheres have been fabricated in the desired size range and equipment is being prepared to establish a uranium dioxide microsphere production capability.

Jeffrey A. Katalenich; Michael R. Hartman; Robert C. O'Brien

2013-02-01T23:59:59.000Z

245

Uranium immobilization and nuclear waste  

SciTech Connect (OSTI)

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.

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

1982-02-01T23:59:59.000Z

246

Vehicle drive module having improved EMI shielding  

DOE Patents [OSTI]

EMI shielding in an electric vehicle drive is provided for power electronics circuits and the like via a direct-mount reference plane support and shielding structure. The thermal support may receive one or more power electronic circuits. The support may aid in removing heat from the circuits through fluid circulating through the support. The support forms a shield from both external EMI/RFI and from interference generated by operation of the power electronic circuits. Features may be provided to permit and enhance connection of the circuitry to external circuitry, such as improved terminal configurations. Modular units may be assembled that may be coupled to electronic circuitry via plug-in arrangements or through interface with a backplane or similar mounting and interconnecting structures.

Beihoff, Bruce C.; Kehl, Dennis L.; Gettelfinger, Lee A.; Kaishian, Steven C.; Phillips, Mark G.; Radosevich, Lawrence D.

2006-11-28T23:59:59.000Z

247

Power converter having improved EMI shielding  

DOE Patents [OSTI]

EMI shielding is provided for power electronics circuits and the like via a direct-mount reference plane support and shielding structure. The thermal support may receive one or more power electronic circuits. The support may aid in removing heat from the circuits through fluid circulating through the support. The support forms a shield from both external EMI/RFI and from interference generated by operation of the power electronic circuits. Features may be provided to permit and enhance connection of the circuitry to external circuitry, such as improved terminal configurations. Modular units may be assembled that may be coupled to electronic circuitry via plug-in arrangements or through interface with a backplane or similar mounting and interconnecting structures.

Beihoff, Bruce C.; Kehl, Dennis L.; Gettelfinger, Lee A.; Kaishian, Steven C.; Phillips, Mark G.; Radosevich, Lawrence D.

2006-06-13T23:59:59.000Z

248

Preserving Ultra-Pure Uranium-233  

SciTech Connect (OSTI)

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 certified reference materials (CRM), all having an isotopic purity of at least 99.4% {sup 233}U and including materials up to 99.996% purity. Current plans include rescuing the purest {sup 233}U materials during a 3-year project beginning in FY 2012 in three phases involving preparations, handling preserved materials, and cleanup. The first year will involve preparations for handling the rescued material for sampling, analysis, distribution, and storage. Such preparations involve modifying or developing work control documents and physical preparations in the laboratory, which include preparing space for new material-handling equipment and procuring and (in some cases) refurbishing equipment needed for handling {sup 233}U or qualifying candidate CRM. Once preparations are complete, an evaluation of readiness will be conducted by independent reviewers to verify that the equipment, work controls, and personnel are ready for operations involving handling radioactive materials with nuclear criticality safety as well as radiological control requirements. The material-handling phase will begin in FY 2013 and be completed early in FY 2014, as currently scheduled. Material handling involves retrieving candidate CRM items from the ORNL storage facility and shipping them to another laboratory at ORNL; receiving and handling rescued items at the laboratory (including any needed initial processing, acquisition and analysis of samples from each item, and preparation for shipment); and shipping bulk material to destination labs or to a yet-to-be-designated storage location. There are seven groups of {sup 233}U identified for handling based on isotopic purity that require the utmost care to prevent cross-contamination. The last phase, cleanup, also will be completed in 2014. It involves cleaning and removing the equipment and material-handling boxes and characterizing, documenting, and disposing of waste. As part of initial planning, the cost of rescuing candidate {sup 233}U items was estimated roughly. The annualized costs were found to be $1,228K in FY 2012, $1,375K in FY 2013,

Krichinsky, Alan M [ORNL; Goldberg, Dr. Steven A. [DOE SC - Chicago Office; Hutcheon, Dr. Ian D. [Lawrence Livermore National Laboratory (LLNL)

2011-10-01T23:59:59.000Z

249

Corrosion-resistant uranium  

DOE Patents [OSTI]

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.

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

1981-10-21T23:59:59.000Z

250

Corrosion-resistant uranium  

DOE Patents [OSTI]

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.

Hovis, Jr., Victor M. (Kingston, TN); Pullen, William C. (Knoxville, TN); Kollie, Thomas G. (Oak Ridge, TN); Bell, Richard T. (Knoxville, TN)

1983-01-01T23:59:59.000Z

251

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma. Uranium

252

Hysteresis prediction inside magnetic shields and application  

SciTech Connect (OSTI)

We have developed a simple model that is able to describe and predict hysteresis behavior inside Mumetal magnetic shields, when the shields are submitted to ultra-low frequency (<0.01 Hz) magnetic perturbations with amplitudes lower than 60??T. This predictive model has been implemented in a software to perform an active compensation system. With this compensation the attenuation of longitudinal magnetic fields is increased by two orders of magnitude. The system is now integrated in the cold atom space clock called PHARAO. The clock will fly onboard the International Space Station in the frame of the ACES space mission.

Mori?, Igor [Observatoire de Paris, SYRTE, Avenue Denfert 77, 75014 Paris (France); CNES, Edouard Belin 18, 31400 Toulouse (France); De Graeve, Charles-Marie [SOGETI High Tech, chemin Laporte 3, 31300 Toulouse (France); Grosjean, Olivier [CNES, Edouard Belin 18, 31400 Toulouse (France); Laurent, Philippe [Observatoire de Paris, SYRTE, Avenue Denfert 77, 75014 Paris (France)

2014-07-15T23:59:59.000Z

253

Thermophysical Properties of Heat Resistant Shielding Material  

SciTech Connect (OSTI)

This project was aimed at determining thermal conductivity, specific heat and thermal expansion of a heat resistant shielding material for neutron absorption applications. These data are critical in predicting the structural integrity of the shielding under thermal cycling and mechanical load. The measurements of thermal conductivity and specific heat were conducted in air at five different temperatures (-31 F, 73.4 F, 140 F, 212 F and 302 F). The transient plane source (TPS) method was used in the tests. Thermal expansion tests were conducted using push rod dilatometry over the continuous range from -40 F (-40 C) to 302 F (150 C).

Porter, W.D.

2004-12-15T23:59:59.000Z

254

Investigation of breached depleted UF{sub 6} cylinders  

SciTech Connect (OSTI)

In June 1990, during a three-site inspection of cylinders being used for long-term storage of solid depleted UF{sub 6}, two 14-ton cylinders at Portsmouth, Ohio, were discovered with holes in the barrel section of the cylinders. An investigation team was immediately formed to determine the cause of the failures and their impact on future storage procedures and to recommend corrective actions. Subsequent investigation showed that the failures most probably resulted from mechanical damage that occurred at the time that the cylinders had been placed in the storage yard. In both cylinders evidence pointed to the impact of a lifting lug of an adjacent cylinder near the front stiffening ring, where deflection of the cylinder could occur only by tearing the cylinder. The impacts appear to have punctured the cylinders and thereby set up corrosion processes that greatly extended the openings in the wall and obliterated the original crack. Fortunately, the reaction products formed by this process were relatively protective and prevented any large-scale loss of uranium. The main factors that precipitated the failures were inadequate spacing between cylinders and deviations in the orientations of lifting lugs from their intended horizontal position. After reviewing the causes and effects of the failures, the team`s principal recommendation for remedial action concerned improved cylinder handling and inspection procedures. Design modifications and supplementary mechanical tests were also recommended to improve the cylinder containment integrity during the stacking operation.

DeVan, J.H. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States)

1991-12-31T23:59:59.000Z

255

High loading uranium fuel plate  

DOE Patents [OSTI]

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.

Wiencek, Thomas C. (Bolingbrook, IL); Domagala, Robert F. (Indian Head Park, IL); Thresh, Henry R. (Palos Heights, IL)

1990-01-01T23:59:59.000Z

256

The New MCNP6 Depletion Capability  

SciTech Connect (OSTI)

The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology.

Fensin, Michael Lorne [Los Alamos National Laboratory; James, Michael R. [Los Alamos National Laboratory; Hendricks, John S. [Los Alamos National Laboratory; Goorley, John T. [Los Alamos National Laboratory

2012-06-19T23:59:59.000Z

257

The Tower Shielding Facility: Its glorious past  

SciTech Connect (OSTI)

The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports.

Muckenthaler, F.J.

1997-05-07T23:59:59.000Z

258

Oxygen Abundance Measurements of SHIELD Galaxies  

E-Print Network [OSTI]

We have derived oxygen abundances for 8 galaxies from the Survey of HI in Extremely Low-mass Dwarfs (SHIELD). The SHIELD survey is an ongoing study of very low-mass galaxies, with M$_{\\rm HI}$ between 10$^{6.5}$ and 10$^{7.5}$ M$_{\\odot}$, that were detected by the Arecibo Legacy Fast ALFA (ALFALFA) survey. H$\\alpha$ images from the WIYN 3.5m telescope show that these 8 SHIELD galaxies each possess one or two active star-forming regions which were targeted with long-slit spectral observations using the Mayall 4m telescope at KPNO. We obtained a direct measurement of the electron temperature by detection of the weak [O III] $\\lambda$4363 line in 2 of the HII regions. Oxygen abundances for the other HII regions were estimated using a strong-line method. When the SHIELD galaxies are plotted on a B-band luminosity-metallicity diagram they appear to suggest a slightly shallower slope to the relationship than normally seen. However, that offset is systematically reduced when the near-infrared luminosity is used ins...

Haurberg, Nathalie C; Cannon, John M; Marshall, Melissa V

2015-01-01T23:59:59.000Z

259

Beer and Economic Growth Dr. Martin Shields  

E-Print Network [OSTI]

Beer and Economic Growth Dr. Martin Shields Regional Economics Institute Colorado State University #12;The Idea · Regional economic growth depends, in part, on the ability to sell goods and services) ­ Industry employment is 35 times more concentrated in Larimer County than the US average! #12;Economic

260

Early test facilities and analytic methods for radiation shielding: Proceedings  

SciTech Connect (OSTI)

This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting. The meeting is of special significance since it commemorates the fiftieth anniversary of the first controlled nuclear chain reaction. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting. The paper titles are good indicators of their content and are: (1) The origin of radiation shielding research: The Oak Ridge experience, (2) Shielding research at the hanford site, (3) Aircraft shielding experiments at General Dynamics Fort Worth, 1950-1962, (4) Where have the neutrons gone , a history of the tower shielding facility, (5) History and evolution of buildup factors, (6) Early shielding research at Bettis atomic power laboratory, (7) UK reactor shielding: then and now, (8) A very personal view of the development of radiation shielding theory.

Ingersoll, D.T. (comp.) (Oak Ridge National Lab., TN (United States)); Ingersoll, J.K. (comp.) (Tec-Com, Knoxville, TN (United States))

1992-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Experimental Test of Self-Shielding in VUV Photodissociation...  

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

Experimental Test of Self-Shielding in VUV Photodissociation of CO Experimental Test of Self-Shielding in VUV Photodissociation of CO Print Wednesday, 25 March 2009 00:00 One way...

262

Uranium from seawater  

SciTech Connect (OSTI)

A novel process for recovering uranium from seawater is proposed and some of the critical technical parameters are evaluated. The process, in summary, consists of two different options for contacting adsorbant pellets with seawater without pumping the seawater. It is expected that this will reduce the mass handling requirements, compared to pumped seawater systems, by a factor of approximately 10/sup 5/, which should also result in a large reduction in initial capital investment. Activated carbon, possibly in combination with a small amount of dissolved titanium hydroxide, is expected to be the preferred adsorbant material instead of the commonly assumed titanium hydroxide alone. The activated carbon, after exposure to seawater, can be stripped of uranium with an appropriate eluant (probably an acid) or can be burned for its heating value (possible in a power plant) leaving the uranium further enriched in its ash. The uranium, representing about 1% of the ash, is then a rich ore and would be recovered in a conventional manner. Experimental results have indicated that activated carbon, acting alone, is not adequately effective in adsorbing the uranium from seawater. We measured partition coefficients (concentration ratios) of approximately 10/sup 3/ in seawater instead of the reported values of 10/sup 5/. However, preliminary tests carried out in fresh water show considerable promise for an extraction system that uses a combination of dissolved titanium hydroxide (in minute amounts) which forms an insoluble compound with the uranyl ion, and the insoluble compound then being sorbed out on activated carbon. Such a system showed partition coefficients in excess of 10/sup 5/ in fresh water. However, the system was not tested in seawater.

Gregg, D.; Folkendt, M.

1982-09-21T23:59:59.000Z

263

Neutral depletion and the helicon density limit  

SciTech Connect (OSTI)

It is straightforward to create fully ionized plasmas with modest rf power in a helicon. It is difficult, however, to create plasmas with density >10{sup 20} m{sup ?3}, because neutral depletion leads to a lack of fuel. In order to address this density limit, we present fast (1 MHz), time-resolved measurements of the neutral density at and downstream from the rf antenna in krypton helicon plasmas. At the start of the discharge, the neutral density underneath the antenna is reduced to 1% of its initial value in 15 ?s. The ionization rate inferred from these data implies that the electron temperature near the antenna is much higher than the electron temperature measured downstream. Neutral density measurements made downstream from the antenna show much slower depletion, requiring 14 ms to decrease by a factor of 1/e. Furthermore, the downstream depletion appears to be due to neutral pumping rather than ionization.

Magee, R. M.; Galante, M. E.; Carr, J. Jr.; Lusk, G.; McCarren, D. W.; Scime, E. E. [West Virginia University, Morgantown, West Virginia 26506 (United States)] [West Virginia University, Morgantown, West Virginia 26506 (United States)

2013-12-15T23:59:59.000Z

264

Effective shielding to measure beam current from an ion source  

SciTech Connect (OSTI)

To avoid saturation, beam current transformers must be shielded from solenoid, quad, and RFQ high stray fields. Good understanding of field distribution, shielding materials, and techniques is required. Space availability imposes compact shields along the beam pipe. This paper describes compact effective concatenated magnetic shields for IFMIF-EVEDA LIPAc LEBT and MEBT and for FAIR Proton Linac injector. They protect the ACCT Current Transformers beyond 37 mT radial external fields. Measurements made at Saclay on the SILHI source are presented.

Bayle, H., E-mail: bayle@bergoz.com [Bergoz Instrumentation, Saint-Genis-Pouilly (France); Delferričre, O.; Gobin, R.; Harrault, F.; Marroncle, J.; Senée, F.; Simon, C.; Tuske, O. [CEA, Saclay (France)] [CEA, Saclay (France)

2014-02-15T23:59:59.000Z

265

Measurement of lanthanum and technetium in uranium fuels by inductively coupled plasma atomic emission spectroscopy.  

SciTech Connect (OSTI)

An important parameter in characterizing an irradiated nuclear fuel is determining the amount of uranium fissioned. By determining the amount of uranium fissioned in the fuel a burnup performance parameter can be calculated, and the amount of fission products left in the fuel can be predicted. The quantity of uranium fissioned can be calculated from the amount of lanthanum and technetium present in the fuel. Lanthanum and technetium were measured in irradiated fuel samples using an Inductively Coupled Plasma Atomic Emission Spectroscopy (ICP-AES) instrument and separation equipment located in a shielded glove-box. A discussion of the method, interferences, detection limits, quality control and a comparison to other work will be presented.

Carney, K.; Crane, P.; Cummings, D.; Krsul, J.; McKnight, R.

1999-06-10T23:59:59.000Z

266

Radiation measurements of uranium ingots from the electrometallurgical treatment of spent fuel.  

SciTech Connect (OSTI)

Radiation measurements and gamma spectroscopy analyses were made on numerous uranium ingots produced during the treatment of Experimental Breeder Reactor-II (EBR-II) spent nuclear fuel. The objective of these measurements was to provide background data for shielding concerns and potential process optimization. The uranium ingots resulted from the processing of both driver and blanket fuel by the electrometallurgical treatment process. The observed variation in the measurements was traced to the levels of certain fission product residues that remained in the uranium ingots produced during spent fuel treatment. A minor process change to hold the material at an elevated temperature for a specified length of time was found to significantly reduce concentrations of high-activity fission products and, thus the radiation field.

Westphal, B. R.; Liaw, J. R.; Krsul, J. R.; Maddison, D. W.; Jensen, B. A.

2003-03-24T23:59:59.000Z

267

Engineering of Ferrite-Graphite Composite Media for Microwave Shields  

E-Print Network [OSTI]

Engineering of Ferrite-Graphite Composite Media for Microwave Shields Marina Koledintseva, PoornaAA@mpei.ru Abstract-- An electromagnetic shielding of objects using ferrite-graphite composites is considered- shielding; dielectric base material; ferrite- graphite composite, Maxwell Garnett formulation I

Koledintseva, Marina Y.

268

Method of preparation of uranium nitride  

DOE Patents [OSTI]

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.

Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

2013-07-09T23:59:59.000Z

269

Validation of a Monte Carlo based depletion methodology via High Flux Isotope Reactor HEU post-irradiation examination measurements  

SciTech Connect (OSTI)

The purpose of this study is to validate a Monte Carlo based depletion methodology by comparing calculated post-irradiation uranium isotopic compositions in the fuel elements of the High Flux Isotope Reactor (HFIR) core to values measured using uranium mass-spectrographic analysis. Three fuel plates were analyzed: two from the outer fuel element (OFE) and one from the inner fuel element (IFE). Fuel plates O-111-8, O-350-1, and I-417-24 from outer fuel elements 5-O and 21-O and inner fuel element 49-I, respectively, were selected for examination. Fuel elements 5-O, 21-O, and 49-1 were loaded into HFIR during cycles 4, 16, and 35, respectively (mid to late 1960s). Approximately one year after each of these elements were irradiated, they were transferred to the High Radiation Level Examination Laboratory (HRLEL) where samples from these fuel plates were sectioned and examined via uranium mass-spectrographic analysis. The isotopic composition of each of the samples was used to determine the atomic percent of the uranium isotopes. A Monte Carlo based depletion computer program, ALEPH, which couples the MCNP and ORIGEN codes, was utilized to calculate the nuclide inventory at the end-of-cycle (EOC). A current ALEPH/MCNP input for HFIR fuel cycle 400 was modified to replicate cycles 4, 16, and 35. The control element withdrawal curves and flux trap loadings were revised, as well as the radial zone boundaries and nuclide concentrations in the MCNP model. The calculated EOC uranium isotopic compositions for the analyzed plates were found to be in good agreement with measurements, which reveals that ALEPH/MCNP can accurately calculate burn-up dependent uranium isotopic concentrations for the HFIR core. The spatial power distribution in HFIR changes significantly as irradiation time increases due to control element movement. Accurate calculation of the end-of-life uranium isotopic inventory is a good indicator that the power distribution variation as a function of space and time is accurately calculated, i.e. an integral check. Hence, the time dependent heat generation source terms needed for reactor core thermal hydraulic analysis, if derived from this methodology, have been shown to be accurate for highly enriched uranium (HEU) fuel.

Chandler, David [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL

2010-01-01T23:59:59.000Z

270

THE RIMINI PROTOCOL Oil Depletion Protocol  

E-Print Network [OSTI]

Soaring oil prices have drawn attention to the issue of the relative supply and demand for crude oil1 THE RIMINI PROTOCOL an Oil Depletion Protocol ~ Heading Off Economic Chaos and Political Conflict During the Second Half of the Age of Oil As proposed at the 2003 Pio Manzu Conference

Keeling, Stephen L.

271

Nuclear conflict and ozone depletion Quick summary  

E-Print Network [OSTI]

Nuclear conflict and ozone depletion Quick summary o Regional nuclear war could cause global which traps pollutants o Nuclear weapons cause explosions, which then causes things around the vicinity to start burning, which in turn releases black carbon; it is not the nuclear material or fallout causing

Toohey, Darin W.

272

Commonness, population depletion and conservation biology  

E-Print Network [OSTI]

and alleviate significant depletion events. Priority species Judgements about extinction risk are key drivers to be targets for conservation invest- ment. Indeed, high extinction risk typifies the most iconic species, flagship or indicator species [2­4]), the use of extinction risk to set conservation priorities has

Queensland, University of

273

Method for fabricating uranium foils and uranium alloy foils  

DOE Patents [OSTI]

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.

Hofman, Gerard L. (Downers Grove, IL); Meyer, Mitchell K. (Idaho Falls, ID); Knighton, Gaven C. (Moore, ID); Clark, Curtis R. (Idaho Falls, ID)

2006-09-05T23:59:59.000Z

274

Thermoforming plastic in lead shield construction  

SciTech Connect (OSTI)

Radiation treatments using low energy X-rays or electrons frequently require a final field defining shield to be placed on the patient's skin. A custom made lead cut-out is used to provide a close fit to a particular patient's surface contours. We have developed a procedure which utilizes POLYFORM thermoplastic to obtain a negative mold of the patient instead of the traditional plaster bandage or dental impression gel. The Polyform is softened in warm water, molded carefully over the patient's surface, and is removed when set or hardened, usually within five minutes. Then lead sheet cut-outs can be formed within this negative. For shielding cut-outs requiring thicker lead sheet, a positive is made from dental stone using this Polyform negative. We have found this procedure to be neat, fast and comfortable for both patient and the dosimetrist.

Abrahams, M.E.; Chow, C.H.; Loyd, M.D. (Univ. of Texas Medical Branch, Galveston (USA))

1989-09-01T23:59:59.000Z

275

Electronically shielded solid state charged particle detector  

DOE Patents [OSTI]

An electronically shielded solid state charged particle detector system having enhanced radio frequency interference immunity includes a detector housing with a detector entrance opening for receiving the charged particles. A charged particle detector having an active surface is disposed within the housing. The active surface faces toward the detector entrance opening for providing electrical signals representative of the received charged particles when the received charged particles are applied to the active surface. A conductive layer is disposed upon the active surface. In a preferred embodiment, a nonconductive layer is disposed between the conductive layer and the active surface. The conductive layer is electrically coupled to the detector housing to provide a substantially continuous conductive electrical shield surrounding the active surface. The inner surface of the detector housing is supplemented with a radio frequency absorbing material such as ferrite. 1 fig.

Balmer, D.K.; Haverty, T.W.; Nordin, C.W.; Tyree, W.H.

1996-08-20T23:59:59.000Z

276

Electronically shielded solid state charged particle detector  

DOE Patents [OSTI]

An electronically shielded solid state charged particle detector system having enhanced radio frequency interference immunity includes a detector housing with a detector entrance opening for receiving the charged particles. A charged particle detector having an active surface is disposed within the housing. The active surface faces toward the detector entrance opening for providing electrical signals representative of the received charged particles when the received charged particles are applied to the active surface. A conductive layer is disposed upon the active surface. In a preferred embodiment, a nonconductive layer is disposed between the conductive layer and the active surface. The conductive layer is electrically coupled to the detector housing to provide a substantially continuous conductive electrical shield surrounding the active surface. The inner surface of the detector housing is supplemented with a radio frequency absorbing material such as ferrite.

Balmer, David K. (155 Coral Way, Broomfield, CO 80020); Haverty, Thomas W. (1173 Logan, Northglenn, CO 80233); Nordin, Carl W. (7203 W. 32nd Ave., Wheatridge, CO 80033); Tyree, William H. (1977 Senda Rocosa, Boulder, CO 80303)

1996-08-20T23:59:59.000Z

277

Supplemental heating of deposition tooling shields  

DOE Patents [OSTI]

A method of reducing particle generation from the thin coating deposited on the internal surfaces of a deposition chamber which undergoes temperature variation greater than 100.degree. C. comprising maintaining the temperature variation of the internal surfaces low enough during the process cycle to keep thermal expansion stresses between the coating and the surfaces under 500 MPa. For titanium nitride deposited on stainless steel, this means keeping temperature variations under approximately 70.degree. C. in a chamber that may be heated to over 350.degree. C. during a typical processing operation. Preferably, a supplemental heater is mounted behind the upper shield and controlled by a temperature sensitive element which provides feedback control based on the temperature of the upper shield.

Ohlhausen, James A. (Albuquerque, NM); Peebles, Diane E. (Albuquerque, NM); Hunter, John A. (Albuquerque, NM); Eckelmeyer, Kenneth H. (Albuquerque, NM)

2000-01-01T23:59:59.000Z

278

Transparent self-cleaning dust shield  

DOE Patents [OSTI]

A transparent electromagnetic shield to protect solar panels and the like from dust deposition. The shield is a panel of clear non-conducting (dielectric) material with embedded parallel electrodes. The panel is coated with a semiconducting film. Desirably the electrodes are transparent. The electrodes are connected to a single-phase AC signal or to a multi-phase AC signal that produces a travelling electromagnetic wave. The electromagnetic field produced by the electrodes lifts dust particles away from the shield and repels charged particles. Deposited dust particles are removed when the electrodes are activated, regardless of the resistivity of the dust. Electrostatic charges on the panel are discharged by the semiconducting film. When used in conjunction with photovoltaic cells, the power for the device may be obtained from the cells themselves. For other surfaces, such as windshields, optical windows and the like, the power must be derived from an external source. One embodiment of the invention employs monitoring and detection devices to determine when the level of obscuration of the screen by dust has reached a threshold level requiring activation of the dust removal feature.

Mazumder, Malay K.; Sims, Robert A.; Wilson, James D.

2005-06-28T23:59:59.000Z

279

GRAVITATIONAL FIELD SHIELDING AND SUPERNOVA EXPLOSIONS  

SciTech Connect (OSTI)

A new mechanism for supernova explosions called gravitational field shielding is proposed, in accord with a five-dimensional fully covariant Kaluza-Klein theory with a scalar field that unifies the four-dimensional Einsteinian general relativity and Maxwellian electromagnetic theory. It is shown that a dense compact collapsing core of a star will suddenly turn off or completely shield its gravitational field when the core collapses to a critical density, which is inversely proportional to the square of mass of the core. As the core suddenly turns off its gravity, the extremely large pressure immediately stops the core collapse and pushes the mantle material of supernova moving outward. The work done by the pressure in the expansion can be the order of energy released in a supernova explosion. The gravity will resume and stop the core from a further expansion when the core density becomes less than the critical density. Therefore, the gravitational field shielding leads a supernova to impulsively explode and form a compact object such as a neutron star as a remnant. It works such that a compressed spring will shoot the oscillator out when the compressed force is suddenly removed.

Zhang, T. X. [Physics Department, Alabama A and M University, Normal, AL 35762 (United States)

2010-12-20T23:59:59.000Z

280

Mechanical Behavior Studies of Depleted Uranium in the Presence of Hydrides  

SciTech Connect (OSTI)

This project addresses critical issues related to aging in the presence of hydrides (UH{sub 3}) in DU and the subsequent effect on mechanical behavior. Rolled DU specimens with three different hydrogen concentrations and the as-rolled condition were studied. The texture measurements indicate that the hydrogen charging is affecting the initial as-rolled DU microstructure/texture. The macroscopic mechanical behavior suggests the existence of a threshold between the 0 wpmm H and 0.3 wppm H conditions. A VPSC simulation of the macroscopic strain-stress behavior, when taking into account only a texture effect, shows no agreement with the experiment. This suggests that the macroscopic mechanical behavior observed is indeed due to the presence of hydrogen/hydrides in the DU bulk. From the lattice strain variation it can be concluded that the hydrogen is affecting the magnitude and/or the nature of CRSS. The metallography indicates the specimens that underwent the hydrogen charging process, developed large grains and twinning, which were enhanced by the presence of hydrogen. Further studies using electron microscopy and modeling will be conducted to learn about the deformation mechanisms responsible for the observed behavior.

Garlea, E.; Morrell, J. S.; Bridges, R. L.; Powell, G. L.; Brown, d. W.; Sisneros, T. A.; Tome, C. N.; Vogel, S. C.

2011-02-14T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Radiological Conditions in Areas of Kuwait with Residues of Depleted Uranium RADIOLOGICAL ASSESSMENT  

E-Print Network [OSTI]

Under the terms of Article III of its Statute, the IAEA is authorized to establish standards of safety for protection against ionizing radiation and to provide for the application of these standards to peaceful nuclear activities. The regulatory related publications by means of which the IAEA establishes safety standards and measures are issued in the IAEA Safety Standards Series. This series covers nuclear safety, radiation safety, transport safety and waste safety, and also general safety (that is, of relevance in two or more of the four areas), and the categories within it are Safety Fundamentals, Safety Requirements and Safety Guides. Safety Fundamentals (blue lettering) present basic objectives, concepts and principles of safety and protection in the development and application of nuclear energy for peaceful purposes. Safety Requirements (red lettering) establish the requirements that must be met to ensure safety. These requirements, which are expressed as ‘shall ’ statements, are governed by the objectives and principles presented in the Safety Fundamentals. Safety Guides (green lettering) recommend actions, conditions or procedures for meeting safety requirements. Recommendations in Safety Guides are expressed as ‘should ’ statements, with the implication that it is necessary to take the measures recommended or equivalent alternative measures to comply with the requirements. The IAEA’s safety standards are not legally binding on Member States but may be adopted by them, at their own discretion, for use in national regulations in respect of their own activities. The standards are binding on the IAEA in relation to its own operations and on States in relation to operations assisted by the IAEA. Information on the IAEA’s safety standards programme (including editions in languages other than English) is available at the IAEA Internet site www.iaea.org/ns/coordinet

unknown authors

282

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana.Program - Libbyof Energy Project,Statement |Construction andDepartment

283

DOE Announces Transfer of Depleted Uranium to Advance the U.S. National  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |Energy UsageAUDITVehiclesTankless orAChiefAppropriation FY 2012 FYEnergy DOESecurity

284

DOE Seeks Contractor for Depleted Uranium Hexafluoride (DUF6) Operations at  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |Energy UsageAUDITVehiclesTankless orAChiefAppropriation FYG 242.1-1of1.9MOhio and Kentucky

285

Unexpected, Stable Form of Uranium Detected | EMSL  

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

Unexpected, Stable Form of Uranium Detected Unexpected, Stable Form of Uranium Detected Insights on underappreciated reaction could shed light on environmental cleanup options...

286

Improvements in EBR-2 core depletion calculations  

SciTech Connect (OSTI)

The need for accurate core depletion calculations in Experimental Breeder Reactor No. 2 (EBR-2) is discussed. Because of the unique physics characteristics of EBR-2, it is difficult to obtain accurate and computationally efficient multigroup flux predictions. This paper describes the effect of various conventional and higher order schemes for group constant generation and for flux computations; results indicate that higher-order methods are required, particularly in the outer regions (i.e. the radial blanket). A methodology based on Nodal Equivalence Theory (N.E.T.) is developed which allows retention of the accuracy of a higher order solution with the computational efficiency of a few group nodal diffusion solution. The application of this methodology to three-dimensional EBR-2 flux predictions is demonstrated; this improved methodology allows accurate core depletion calculations at reasonable cost. 13 refs., 4 figs., 3 tabs.

Finck, P.J.; Hill, R.N.; Sakamoto, S.

1991-01-01T23:59:59.000Z

287

Carbon sequestration in depleted oil shale deposits  

SciTech Connect (OSTI)

A method and apparatus are described for sequestering carbon dioxide underground by mineralizing the carbon dioxide with coinjected fluids and minerals remaining from the extraction shale oil. In one embodiment, the oil shale of an illite-rich oil shale is heated to pyrolyze the shale underground, and carbon dioxide is provided to the remaining depleted oil shale while at an elevated temperature. Conditions are sufficient to mineralize the carbon dioxide.

Burnham, Alan K; Carroll, Susan A

2014-12-02T23:59:59.000Z

288

Radiation Shielding for Electronic Devices OperatingRadiation Shielding for Electronic Devices Operating in XFEL Environment: Monte Carlo Simulations andin XFEL Environment: Monte Carlo Simulations and  

E-Print Network [OSTI]

Radiation Shielding for Electronic Devices OperatingRadiation Shielding for Electronic Devices undergroundund tunnel. All LLRF Electronic Devices, made of radiation sensitivetunnel. All LLRF Electronic principle of the dedicated radiationtion shielding for the electronic devices to be operating in XFEL

289

2013 Domestic Uranium Production Report  

E-Print Network [OSTI]

Administration (EIA), the statistical and analytical agency within the U.S. Department of Energy. By law, EIA.S. Energy Information Administration | 2013 Domestic Uranium Production Report iii Preface The U.S. Energy://www.eia.doe.gov/glossary/. #12;U.S. Energy Information Administration | 2013 Domestic Uranium Production Report iv Contents

290

Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel  

SciTech Connect (OSTI)

The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

Hermann, O.W.

2000-02-01T23:59:59.000Z

291

DOE/EA-1607: Final Environmental Assessment for Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium (June 2009)  

Office of Environmental Management (EM)

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292

Domestic Uranium Production Report  

Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines AboutDecember 2005 (Thousand9, 2015Year109 AppendixCostsDistributedSep-1410. Uranium

293

Domestic Uranium Production Report  

Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines AboutDecember 2005 (Thousand9, 2015Year109 AppendixCostsDistributedSep-1410. Uranium9.

294

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.

295

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.

296

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.3.

297

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.3.5.

298

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.3.5.3.

299

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from

300

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma. Uraniumb.

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma. Uraniumb.7.

302

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma.

303

Uranium Marketing Annual Report -  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma.9.

304

Fingerprinting Uranium | EMSL  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsing ZirconiaPolicyFeasibilityField Office FinalFinancingFingerprinting Uranium

305

Uranium Enrichment Standards of the Y-12 Nuclear Detection and Sensor Testing Center  

SciTech Connect (OSTI)

The Y-12 National Security Complex has recently fabricated and characterized a new series of metallic uranium standards for use in the Nuclear Detection and Sensor Testing Center (NDSTC). Ten uranium metal disks with enrichments varying from 0.2 to 93.2% {sup 235}U were designed to provide researchers access to a wide variety of measurement scenarios in a single testing venue. Special care was taken in the selection of the enrichments in order to closely bracket the definitions of reactor fuel at 4% {sup 235}U and that of highly enriched uranium (HEU) at 20% {sup 235}U. Each standard is well characterized using analytical chemistry as well as a series of gamma-ray spectrometry measurements. Gamma-ray spectra of these standards are being archived in a reference library for use by customers of the NDSTC. A software database tool has been created that allows for easier access and comparison of various spectra. Information provided through the database includes: raw count data (including background spectra), regions of interest (ROIs), and full width half maximum calculations. Input is being sought from the user community on future needs including enhancements to the spectral database and additional Uranium standards, shielding configurations and detector types. A related presentation are planned for the INMM 53rd Annual Meeting (Hull, et al.), which describe new uranium chemical compound standards and testing opportunities at Y-12 Nuclear Detection and Sensor Testing Center (NDSTC).

Cantrell, J.

2012-05-23T23:59:59.000Z

306

RESEARCH ARTICLE Open Access Susceptibility to ATP depletion of primary  

E-Print Network [OSTI]

RESEARCH ARTICLE Open Access Susceptibility to ATP depletion of primary proximal tubular cell subjected to ATP depletion using antimycin A. Results: Surprisingly, there was no difference in the amount, Viability, Survival, Apoptosis knockout mice, shRNA, ATP depletion, Metabolic stress, Antimycin Background

Paris-Sud XI, Université de

307

Method for making a uranium chloride salt product  

DOE Patents [OSTI]

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.

Miller, William E. (Naperville, IL); Tomczuk, Zygmunt (Lockport, IL)

2004-10-05T23:59:59.000Z

308

Validity of Hansen-Roach cross sections in low-enriched uranium systems  

SciTech Connect (OSTI)

Within the nuclear criticality safety community, the Hansen-Roach 16 group cross section set has been the standard'' for use in k{sub eff} calculations over the past 30 years. Yet even with its widespread acceptance, there are still questions about its validity and adequacy, about the proper procedure for calculating the potential scattering cross section, {sigma}{sub p}, for uranium and plutonium, and about the concept of resonance self shielding and its impact on cross sections. This paper attempts to address these questions. It provides a brief background on the Hansen-Roach cross sections. Next is presented a review of resonances in cross sections, self shielding of these resonances, and the use of {sigma}{sub p} to characterize resonance self shielding. Three prescriptions for calculating {sigma}{sub p} are given. Finally, results of several calculations of k{sub eff} on low-enriched uranium systems are provided to confirm the validity of the Hansen-Roach cross sections when applied to such systems.

Busch, R.D. (New Mexico Univ., Albuquerque, NM (United States)); O'Dell, R.D. (Los Alamos National Lab., NM (United States))

1991-01-01T23:59:59.000Z

309

Cross Section Evaluation Group shielding benchmark compilation. Volume II  

SciTech Connect (OSTI)

At the time of the release of ENDF/B-IV in 1974, the Shielding Subcommittee had identified a series of 12 shielding data testing benchmarks (the SDT series). Most were used in the ENDF/B-IV data testing effort. A new concept and series was begun in the interim, the so-called Shielding Benchmark (SB) series. An effort was made to upgrade the SDT series as far as possible and to add new SB benchmarks. In order to be designated in the SB class, both an experiment and analysis must have been performed. The current recommended benchmark for Shielding Data Testing are listed. Until recently, the philosophy has been to include only citations to published references for shielding benchmarks. It is now our intention to provide adequate information in this volume for proper analysis of any new benchmarks added to the collection. These compilations appear in Section II, with the SB5 Fusion Reactor Shielding Benchmark as the first entry.

Rose, P.F.; Roussin, R.W.

1983-12-01T23:59:59.000Z

310

Radiation shielding for underground low-background experiments  

E-Print Network [OSTI]

The design task of creating an efficient radiation shield for the new COBRA double-beta decay experiment led to a comprehensive study of commercially available shielding materials. The aim was to find the most efficient combination of materials under the constraints of an extreme low-background experiment operating in a typical underground laboratory. All existing shield configurations for this type of experiment have been found to perform sub-optimally in comparison to the class of multilayered configurations proposed in this study. The method used here to create a specific shield configuration should yield a close to optimal result when applied to any experiment utilising a radiation shield. In particular, the survey of single material response to a given radiation source turns out to give a guideline for the construction of efficient multilayer shields.

D. Y Stewart; P. F. Harrison; B. Morgan; Y. A. Ramachers

2006-07-31T23:59:59.000Z

311

Shielded serpentine traveling wave tube deflection structure  

DOE Patents [OSTI]

A shielded serpentine slow wave deflection structure (10) having a serpene signal conductor (12) within a channel groove (46). The channel groove (46) is formed by a serpentine channel (20) in a trough plate (18) and a ground plane (14). The serpentine signal conductor (12) is supported at its ends by coaxial feed through connectors 28. A beam interaction trough (22) intersects the channel groove (46) to form a plurality of beam interaction regions (56) wherein an electron beam (54) may be deflected relative to the serpentine signal conductor (12).

Hudson, Charles L. (Santa Barbara, CA); Spector, Jerome (Berkeley, CA)

1994-01-01T23:59:59.000Z

312

Longwall shield design: is bigger better?  

SciTech Connect (OSTI)

This article evaluates the bigger is better design philosophy for longwall shields. The conventional support design approach based on simplistic models of supporting the full dead weight detached rock masses is replaced by a ground reaction design approach. Here, the goal is to match the support characteristics to the ground response, and not to try and overpower the ground forces with some massive support capability. The ground reaction concept embodies both the force and displacement controlled loading aspects, and therefore provides a more accurate representation of the support loading requirements. 7 figs.

Barczak, T.M.; Tadolini, S.C. [NIOSH-PRL, Pittsburgh, PA (United States)

2008-05-15T23:59:59.000Z

313

X-ray transmissive debris shield  

DOE Patents [OSTI]

An X-ray debris shield for use in X-ray lithography that is comprised of an X-ray window having a layer of low density foam exhibits increased longevity without a substantial increase in exposure time. The low density foam layer serves to absorb the debris emitted from the X-ray source and attenuate the shock to the window so as to reduce the chance of breakage. Because the foam is low density, the X-rays are hardly attenuated by the foam and thus the exposure time is not substantially increased.

Spielman, Rick B. (Albuquerque, NM)

1996-01-01T23:59:59.000Z

314

APPENDIX J Partition Coefficients For Uranium  

E-Print Network [OSTI]

APPENDIX J Partition Coefficients For Uranium #12;Appendix J Partition Coefficients For Uranium J.1.0 Background The review of uranium Kd values obtained for a number of soils, crushed rock and their effects on uranium adsorption on soils are discussed below. The solution pH was also used as the basis

315

Methods and Procedures for Shielding Analyses for the SNS  

SciTech Connect (OSTI)

In order to provide radiologically safe Spallation Neutron Source operation, shielding analyses are performed according to Oak Ridge National Laboratory internal regulations and to comply with the Code of Federal Regulations. An overview of on-going shielding work for the accelerator facility and neutrons beam lines, methods, used for the analyses, and associated procedures and regulations is presented. Methods used to perform shielding analyses are described as well.

Gallmeier, Franz X [ORNL] [ORNL; Iverson, Erik B [ORNL] [ORNL; Remec, Igor [ORNL] [ORNL; Lu, Wei [ORNL] [ORNL; Popova, Irina [ORNL] [ORNL

2014-01-01T23:59:59.000Z

316

Electromagnetic Interference (EMI) Shielding of Single-Walled Carbon  

E-Print Network [OSTI]

Electromagnetic Interference (EMI) Shielding of Single-Walled Carbon Nanotube Epoxy Composites Ning (SWNT)-polymer composites have been fabricated to evaluate the electromagnetic interference (EMI

Gao, Hongjun

317

Gravity Scaling of a Power Reactor Water Shield  

SciTech Connect (OSTI)

Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa{sup n}. These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

Reid, Robert S.; Pearson, J. Boise [NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States)

2008-01-21T23:59:59.000Z

318

Graphene shield enhanced photocathodes and methods for making the same  

SciTech Connect (OSTI)

Disclosed are graphene shield enhanced photocathodes, such as high QE photocathodes. In certain embodiments, a monolayer graphene shield membrane ruggedizes a high quantum efficiency photoemission electron source by protecting a photosensitive film of the photocathode, extending operational lifetime and simplifying its integration in practical electron sources. In certain embodiments of the disclosed graphene shield enhanced photocathodes, the graphene serves as a transparent shield that does not inhibit photon or electron transmission but isolates the photosensitive film of the photocathode from reactive gas species, preventing contamination and yielding longer lifetime.

Moody, Nathan Andrew

2014-09-02T23:59:59.000Z

319

The End of Cheap Uranium  

E-Print Network [OSTI]

Historic data from many countries demonstrate that on average no more than 50-70% of the uranium in a deposit could be mined. An analysis of more recent data from Canada and Australia leads to a mining model with an average deposit extraction lifetime of 10+- 2 years. This simple model provides an accurate description of the extractable amount of uranium for the recent mining operations. Using this model for all larger existing and planned uranium mines up to 2030, a global uranium mining peak of at most 58 +- 4 ktons around the year 2015 is obtained. Thereafter we predict that uranium mine production will decline to at most 54 +- 5 ktons by 2025 and, with the decline steepening, to at most 41 +- 5 ktons around 2030. This amount will not be sufficient to fuel the existing and planned nuclear power plants during the next 10-20 years. In fact, we find that it will be difficult to avoid supply shortages even under a slow 1%/year worldwide nuclear energy phase-out scenario up to 2025. We thus suggest that a worldwide nuclear energy phase-out is in order. If such a slow global phase-out is not voluntarily effected, the end of the present cheap uranium supply situation will be unavoidable. The result will be that some countries will simply be unable to afford sufficient uranium fuel at that point, which implies involuntary and perhaps chaotic nuclear phase-outs in those countries involving brownouts, blackouts, and worse.

Michael Dittmar

2011-06-21T23:59:59.000Z

320

MicroShield/ISOCS gamma modeling comparison.  

SciTech Connect (OSTI)

Quantitative radiological analysis attempts to determine the quantity of activity or concentration of specific radionuclide(s) in a sample. Based upon the certified standards that are used to calibrate gamma spectral detectors, geometric similarities between sample shape and the calibration standards determine if the analysis results developed are qualitative or quantitative. A sample analyzed that does not mimic a calibrated sample geometry must be reported as a non-standard geometry and thus the results are considered qualitative and not quantitative. MicroShieldR or ISOCSR calibration software can be used to model non-standard geometric sample shapes in an effort to obtain a quantitative analytical result. MicroShieldR and Canberra's ISOCSR software contain several geometry templates that can provide accurate quantitative modeling for a variety of sample configurations. Included in the software are computational algorithms that are used to develop and calculate energy efficiency values for the modeled sample geometry which can then be used with conventional analysis methodology to calculate the result. The response of the analytical method and the sensitivity of the mechanical and electronic equipment to the radionuclide of interest must be calibrated, or standardized, using a calibrated radiological source that contains a known and certified amount of activity.

Sansone, Kenneth R

2013-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

SHIELDING ANALYSIS FOR PORTABLE GAUGING COMBINATION SOURCES  

SciTech Connect (OSTI)

Radioisotopic decay has been used as a source of photons and neutrons for industrial gauging operations since the late 1950s. Early portable moisture/density gauging equipment used Americium (Am)-241/Beryllium (Be)/Cesium (Cs)-137 combination sources to supply the required nuclear energy for gauging. Combination sources typically contained 0.040 Ci of Am-241 and 0.010 Ci of CS-137 in the same source capsule. Most of these sources were manufactured approximately 30 years ago. Collection, transportation, and storage of these sources once removed from their original device represent a shielding problem with distinct gamma and neutron components. The Off-Site Source Recovery (OSR) Project is planning to use a multi-function drum (MFD) for the collection, shipping, and storage of AmBe sources, as well as the eventual waste package for disposal. The MFD is an approved TRU waste container design for DOE TRU waste known as the 12 inch Pipe Component Overpack. As the name indicates, this drum is based on a 12 inch ID stainless steel weldment approximately 25 inch in internal length. The existing drum design allows for addition of shielding within the pipe component up to the 110 kg maximum pay load weight. The 12 inch pipe component is packaged inside a 55-gallon drum, with the balance of the interior space filled with fiberboard dunnage. This packaging geometry is similar to the design of a DOT 6M, Type B shipping container.

J. TOMPKINS; L. LEONARD; ET AL

2000-08-01T23:59:59.000Z

322

System for imaging plutonium through heavy shielding  

SciTech Connect (OSTI)

A single pinhole can be used to image strong self-luminescent gamma-ray sources such as plutonium on gamma scintillation (Anger) cameras. However, if the source is weak or heavily shielded, a poor signal to noise ratio can prevent acquisition of the image. An imaging system designed and built at Los Alamos National Laboratory uses a coded aperture to image heavily shielded sources. The paper summarizes the mathematical techniques, based on the Fast Delta Hadamard transform, used to decode raw images. Practical design considerations such as the phase of the uniformly redundant aperture and the encoded image sampling are discussed. The imaging system consists of a custom designed m-sequence coded aperture, a Picker International Corporation gamma scintillation camera, a LeCroy 3500 data acquisition system, and custom imaging software. The paper considers two sources - 1.5 mCi /sup 57/Co unshielded at a distance of 27 m and 220 g of bulk plutonium (11.8% /sup 240/Pu) with 0.3 cm lead, 2.5 cm steel, and 10 cm of dense plastic material at a distance of 77.5 cm. Results show that the location and geometry of a source hidden in a large sealed package can be determined without having to open the package. 6 references, 4 figures.

Kuckertz, T.H.; Cannon, T.M.; Fenimore, E.E.; Moss, C.E.; Nixon, K.V.

1984-04-01T23:59:59.000Z

323

Liquid Vortex Shielding for Fusion Energy Applications  

SciTech Connect (OSTI)

Swirling liquid vortices can be used in fusion chambers to protect their first walls and critical elements from the harmful conditions resulting from fusion reactions. The beam tube structures in heavy ion fusion (HIF) must be shielded from high energy particles, such as neutrons, x-rays and vaporized coolant, that will cause damage. Here an annular wall jet, or vortex tube, is proposed for shielding and is generated by injecting liquid tangent to the inner surface of the tube both azimuthally and axially. Its effectiveness is closely related to the vortex tube flow properties. 3-D particle image velocimetry (PIV) is being conducted to precisely characterize its turbulent structure. The concept of annular vortex flow can be extended to a larger scale to serve as a liquid blanket for other inertial fusion and even magnetic fusion systems. For this purpose a periodic arrangement of injection and suction holes around the chamber circumference are used, generating the layer. Because it is important to match the index of refraction of the fluid with the tube material for optical measurement like PIV, a low viscosity mineral oil was identified and used that can also be employed to do scaled experiments of molten salts at high temperature.

Bardet, Philippe M. [University of California, Berkeley (United States); Supiot, Boris F. [University of California, Berkeley (United States); Peterson, Per F. [University of California, Berkeley (United States); Savas, Oemer [University of California, Berkeley (United States)

2005-05-15T23:59:59.000Z

324

DEPARTMENT OF ENERGY Excess Uranium Management: Effects of DOE...  

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

Excess Uranium Management: Effects of DOE Transfers of Excess Uranium on Domestic Uranium Mining, Conversion, and Enrichment Industries; Request for Information AGENCY: Office of...

325

General Corrosion and Localized Corrosion of the Drip Shield  

SciTech Connect (OSTI)

The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall. The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared according to ''Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The models developed in this report are used by the waste package degradation analyses for TSPA-LA and serve as a basis to determine the performance of the drip shield. The drip shield may suffer from other forms of failure such as the hydrogen induced cracking (HIC) or stress corrosion cracking (SCC), or both. Stress corrosion cracking of the drip shield material is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]). Hydrogen induced cracking of the drip shield material is discussed in ''Hydrogen Induced Cracking of Drip Shield'' (BSC 2004 [DIRS 169847]).

F. Hua

2004-09-16T23:59:59.000Z

326

Issues and test requirements in radiation shielding of fusion reactors  

SciTech Connect (OSTI)

Radiation shield issues for fusion reactors have been investigated and the experiments and facilities required to resolve the issues have been identified and characterized as part of the FINESSE program. This paper summarizes the recommended approach to fusion shield research and development, provides a summary of the necessary experiments and facilities, and presents the results of technical analyses involved.

Nakagawa, M.; Abdou, M.A.

1986-11-01T23:59:59.000Z

327

Issues and test requirements in radiation shielding of fusion reactors  

SciTech Connect (OSTI)

Radiation shield issues for fusion reactors have been investigated and the experiments and facilities required to resolve the issues have been identified and characterized as part of the FINESSE program. This paper summarizes the recommended approach to fusion shield R and D, provides a summary of the necessary experiments and facilities, and presents the results of technical analyses involved.

Nakagawa, M.; Abdou, M.A.

1986-01-01T23:59:59.000Z

328

Laser induced phosphorescence uranium analysis  

DOE Patents [OSTI]

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.

Bushaw, Bruce A. (Kennewick, WA)

1986-01-01T23:59:59.000Z

329

Laser induced phosphorescence uranium analysis  

DOE Patents [OSTI]

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.

Bushaw, B.A.

1983-06-10T23:59:59.000Z

330

An Optically Stimulated Luminescence Uranium Enrichment Monitor  

SciTech Connect (OSTI)

The Pacific Northwest National Laboratory (PNNL) has pioneered the use of Optically Stimulated Luminescence (OSL) technology for use in personnel dosimetry and high dose radiation processing dosimetry. PNNL has developed and patented an alumina-based OSL dosimeter that is being used by the majority of medical X-ray and imaging technicians worldwide. PNNL has conceived of using OSL technology to passively measure the level of UF6 enrichment by attaching the prototype OSL monitor to pipes containing UF6 gas within an enrichment facility. The prototype OSL UF6 monitor utilizes a two-element approach with the first element open and unfiltered to measure both the low energy and high energy gammas from the UF6, while the second element utilizes a 3-mm thick tungsten filter to eliminate the low energy gammas and pass only the high energy gammas from the UF6. By placing a control monitor in the room away from the UF6 pipes and other ionizing radiation sources, the control readings can be subtracted from the UF6 pipe monitor measurements. The ratio of the shielded to the unshielded net measurements provides a means to estimate the level of uranium enrichment. PNNL has replaced the commercially available MicroStar alumina-based dosimeter elements with a composite of polyethylene plastic, high-Z glass powder, and BaFBr:Eu OSL phosphor powder at various concentrations. The high-Z glass was added in an attempt to raise the average “Z” of the composite dosimeter and increase the response. Additionally, since BaFBr:Eu OSL phosphor is optimally excited and emits light at different wavelengths compared to alumina, the commercially available MicroStar reader was modified for reading BaFBr:Eu in a parallel effort to increase reader sensitivity. PNNL will present the design and performance of our novel OSL uranium enrichment monitor based on a combination of laboratory and UF6 test loop measurements. PNNL will also report on the optimization effort to achieve the highest possible performance from both the OSL enrichment monitor and the new custom OSL reader modified for this application. This project has been supported by the US Department of Energy’s National Nuclear Security Administration’s Office of Dismantlement and Transparency (DOE/NNSA/NA-241).

Miller, Steven D.; Tanner, Jennifer E.; Simmons, Kevin L.; Conrady, Matthew M.; Benz, Jacob M.; Greenfield, Bryce A.

2010-08-11T23:59:59.000Z

331

Fan-fold shielded electrical leads  

DOE Patents [OSTI]

Disclosed are fan-folded electrical leads made from copper cladded Kapton, for example, with the copper cladding on one side serving as a ground plane and the copper cladding on the other side being etched to form the leads. The Kapton is fan folded with the leads located at the bottom of the fan-folds. Electrical connections are made by partially opening the folds of the fan and soldering, for example, the connections directly to the ground plane and/or the lead. The fan folded arrangement produces a number of advantages, such as electrically shielding the leads from the environment, is totally non-magnetic, and has a very low thermal conductivity, while being easy to fabricate. 3 figs.

Rohatgi, R.R.; Cowan, T.E.

1996-06-11T23:59:59.000Z

332

Fan-fold shielded electrical leads  

DOE Patents [OSTI]

Fan-folded electrical leads made from copper cladded Kapton, for example, with the copper cladding on one side serving as a ground plane and the copper cladding on the other side being etched to form the leads. The Kapton is fan folded with the leads located at the bottom of the fan-folds. Electrical connections are made by partially opening the folds of the fan and soldering, for example, the connections directly to the ground plane and/or the lead. The fan folded arrangement produces a number of advantages, such as electrically shielding the leads from the environment, is totally non-magnetic, and has a very low thermal conductivity, while being easy to fabricate.

Rohatgi, Rajeev R. (Mountain View, CA); Cowan, Thomas E. (Livermore, CA)

1996-01-01T23:59:59.000Z

333

Protective shield for an instrument probe  

DOE Patents [OSTI]

A shield is disclosed that is particularly useful for protecting exposed optical elements at the end of optical probes used in the analysis of hazardous emissions in and around an industrial environment from the contaminating effects of those emissions. The instant invention provides a hood or cowl in the shape of a right circular cylinder that can be fitted over the end of such optical probes. The hood provides a clear aperture through which the probe can perform unobstructed analysis. The probe optical elements are protected from the external environment by passing a dry gas through the interior of the hood and out through the hood aperture in sufficient quantity and velocity to prevent any significant mixing between the internal and external environments. Additionally, the hood is provided with a cooling jacket to lessen the potential for damaging the probe due to temperature excursions.

Johnsen, Howard A.; Ross, James R.; Birtola, Sal R.

2004-10-26T23:59:59.000Z

334

Standard Test Method for Determination of Uranium, Oxygen to Uranium (O/U), and Oxygen to Metal (O/M) in Sintered Uranium Dioxide and Gadolinia-Uranium Dioxide Pellets by Atmospheric Equilibration  

E-Print Network [OSTI]

Standard Test Method for Determination of Uranium, Oxygen to Uranium (O/U), and Oxygen to Metal (O/M) in Sintered Uranium Dioxide and Gadolinia-Uranium Dioxide Pellets by Atmospheric Equilibration

American Society for Testing and Materials. Philadelphia

2007-01-01T23:59:59.000Z

335

Geology and recognition criteria for uranium deposits of the quartz-pebble conglomerate type. Final report  

SciTech Connect (OSTI)

This report is concerned with Precambrian uraniferous conglomerates. This class of deposit has been estimated to contain between approximately 16 and 35 percent of the global uranium reserve in two rather small areas, one in Canada, the other in South Africa. Similar conglomerates, which are often gold-bearing, are, however, rather widespread, being found in parts of most Precambrian shield areas. Data have been synthesized on the geologic habitat and character of this deposit type. The primary objective has been to provide the most relevant geologic observations in a structural fashion to allow resource studies and exploration to focus on the most prospective targets in the shortest possible time.

Button, A.; Adams, S.S.

1981-03-01T23:59:59.000Z

336

Lithium Depletion of Nearby Young Stellar Associations  

E-Print Network [OSTI]

We estimate cluster ages from lithium depletion in five pre-main-sequence groups found within 100 pc of the Sun: TW Hydrae Association, Eta Chamaeleontis Cluster, Beta Pictoris Moving Group, Tucanae-Horologium Association and AB Doradus Moving Group. We determine surface gravities, effective temperatures and lithium abundances for over 900 spectra through least squares fitting to model-atmosphere spectra. For each group, we compare the dependence of lithium abundance on temperature with isochrones from pre-main-sequence evolutionary tracks to obtain model dependent ages. We find that the Eta Chamaelontis Cluster and the TW Hydrae Association are the youngest, with ages of 12+/-6 Myr and 12+/-8 Myr, respectively, followed by the Beta Pictoris Moving Group at 21+/-9 Myr, the Tucanae-Horologium Association at 27+/-11 Myr, and the AB Doradus Moving Group at an age of at least 45 Myr (where we can only set a lower limit since the models -- unlike real stars -- do not show much lithium depletion beyond this age). Here, the ordering is robust, but the precise ages depend on our choice of both atmospheric and evolutionary models. As a result, while our ages are consistent with estimates based on Hertzsprung-Russell isochrone fitting and dynamical expansion, they are not yet more precise. Our observations do show that with improved models, much stronger constraints should be feasible: the intrinsic uncertainties, as measured from the scatter between measurements from different spectra of the same star, are very low: around 10 K in effective temperature, 0.05 dex in surface gravity, and 0.03 dex in lithium abundance.

Erin Mentuch; Alexis Brandeker; Marten H. van Kerkwijk; Ray Jayawardhana; Peter H. Hauschildt

2008-08-26T23:59:59.000Z

337

Deuterium depletion and magnesium enhancement in the local disc  

E-Print Network [OSTI]

The local disc deuter is known to be depleted in comparison to the local bubble. We show, that the same lines of sight that are depleted in deuter, are enhanced in magnesium. Heavier elements - Si and Fe do not show any difference in the abundance between the local disc and the local bubble. This observation implicates that astration is responsible for both deuter depletion and magnesium enhancement.

Piotr Gnacinski

2005-07-19T23:59:59.000Z

338

E-Print Network 3.0 - analytical shielding calculations Sample...  

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

Use of Magnetic and Electrically Conductive Fillers in a Polymer Matrix for Electromagnetic Interference Shielding Summary: for Electromagnetic Interference Shielding JUNHUA WU1,2...

339

Polyethylene as a Radiation Shielding Standard in Simulated Cosmic-Ray Environments  

E-Print Network [OSTI]

on the ISS through polyethylene shielding augmentation ofnucleon Iron-56 in Polyethylene. II. , Comparisons betweenPolyethylene as a Radiation Shielding Standard in Simulated

2006-01-01T23:59:59.000Z

340

RELATIVE ATTENUATION CHARACTERISTICS OF SOME SHIELDING MATERIALS FOR PuB NEUTRONS  

E-Print Network [OSTI]

1: Polyethylene Water Spodumene-gypsum Gypsum, wet and dryconstituents of the spodumene-gypsum, and gypsum shields.SPODUMENK·,GYPSUM SHIELD 30% Spodumene by weight 40% Gypsum

Bringham, P.S.

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

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

SciTech Connect (OSTI)

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

Barlow, C.R.; Alderson, J.H.; Blue, S.C.; Boelens, R.A.; Conkel, M.E.; Dorning, R.E.; Ecklund, C.D.; Halicks, W.G.; Henson, H.M.; Newman, V.S.; Philpot, H.E.; Taylor, M.S.; Vournazos, J.P. [Oak Ridge K-25 Site, TN (United States). UEO Enrichment Technical Operations Div.; Russell, J.R. [USDOE Oak Ridge Field Office, TN (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States); Ziehlke, K.T. [MJB Technical Associates (United States)

1992-07-01T23:59:59.000Z

342

Analysis of Serum Total and Free PSA Using Immunoaffinity Depletion...  

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

Serum Total and Free PSA Using Immunoaffinity Depletion Coupled to SRM: Correlation with Clinical Immunoassay Tests. Analysis of Serum Total and Free PSA Using Immunoaffinity...

343

Application of thermal depletion model to geothermal reservoirs...  

Open Energy Info (EERE)

of thermal depletion model to geothermal reservoirs with fracture and pore permeability Jump to: navigation, search OpenEI Reference LibraryAdd to library Conference Proceedings:...

344

Microscale Depletion of High Abundance Proteins in Human Biofluids...  

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

by nonspecific binding to the column matrix. Additionally, the cost of the depletion media can be prohibitive for larger scale studies. Modern LC-MS instrumentation provides...

345

Evidence of uranium biomineralization in sandstone-hosted roll-front uranium deposits, northwestern China  

E-Print Network [OSTI]

Evidence of uranium biomineralization in sandstone-hosted roll-front uranium deposits, northwestern Available online 25 January 2005 Abstract We show evidence that the primary uranium minerals, uraninite-front uranium deposits, Xinjiang, northwestern China were biogenically precipitated and psuedomorphically

Fayek, Mostafa

346

Inherently safe in situ uranium recovery  

DOE Patents [OSTI]

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.

Krumhansl, James L; Brady, Patrick V

2014-04-29T23:59:59.000Z

347

The End of Cheap Uranium  

E-Print Network [OSTI]

Historic data from many countries demonstrate that on average no more than 50-70% of the uranium in a deposit could be mined. An analysis of more recent data from Canada and Australia leads to a mining model with an average deposit extraction lifetime of 10+- 2 years. This simple model provides an accurate description of the extractable amount of uranium for the recent mining operations. Using this model for all larger existing and planned uranium mines up to 2030, a global uranium mining peak of at most 58 +- 4 ktons around the year 2015 is obtained. Thereafter we predict that uranium mine production will decline to at most 54 +- 5 ktons by 2025 and, with the decline steepening, to at most 41 +- 5 ktons around 2030. This amount will not be sufficient to fuel the existing and planned nuclear power plants during the next 10-20 years. In fact, we find that it will be difficult to avoid supply shortages even under a slow 1%/year worldwide nuclear energy phase-out scenario up to 2025. We thus suggest that a world...

Dittmar, Michael

2011-01-01T23:59:59.000Z

348

Shielding design for the proposed Advanced Photon Source at Argonne  

SciTech Connect (OSTI)

Bulk shielding was designed for the proposed Argonne Advanced Photon Source. The shielding is for two linacs, the positron converter, booster synchrotron, and the storage ring. Shielding design limits exposure to 20 mrem/wk for occupational and 25 mrem/y for an individual member of the public from the radiation products, which include high energy neutrons (HEN), giant resonance neutrons (GRN), and Bremsstrahlung radiation (BR). The beam loss parameters at various components were estimated. Dose rates were computed for continuous loss during beam decay using an empirical method. Normal operational losses and certain accidental beam losses were also considered.

Moe, H.J.; Veluri, V.R.

1987-01-01T23:59:59.000Z

349

Radiolysis Concerns for Water Shielding in Fission Surface Power Applications  

SciTech Connect (OSTI)

This paper presents an overview of radiolysis concerns with regard to water shields for fission surface power. A review of the radiolysis process is presented and key parameters and trends are identified. From this understanding of the radiolytic decomposition of water, shield pressurization and corrosion are identified as the primary concerns. Existing experimental and modeling data addressing concerns are summarized. It was found that radiolysis of pure water in a closed volume results in minimal, if any net decomposition, and therefore reduces the potential for shield pressurization and corrosion.

Schoenfeld, Michael P. [NASA Marshall Space Flight Center, ER24, MSFC, AL 35812 (United States); Anghaie, Samim [Innovative Space Power and Propulsion Institute, 800 SW Archer Rd. Bldg.554, P.O. Box 116502, University of Florida, Gainesville, FL 32611-6502 (United States)

2008-01-21T23:59:59.000Z

350

Conversion and Blending Facility highly enriched uranium to low enriched uranium as uranyl nitrate hexahydrate. Revision 1  

SciTech Connect (OSTI)

This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials to pure HEU uranyl nitrate (UNH) and (2) blend pure HEU UNH with depleted and natural UNH to produce HEU UNH crystals. 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. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU Will be produced as a waste suitable for storage or disposal.

NONE

1995-07-05T23:59:59.000Z

351

High strength uranium-tungsten alloys  

DOE Patents [OSTI]

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.

Dunn, Paul S. (Santa Fe, NM); Sheinberg, Haskell (Los Alamos, NM); Hogan, Billy M. (Los Alamos, NM); Lewis, Homer D. (Bayfield, CO); Dickinson, James M. (Los Alamos, NM)

1991-01-01T23:59:59.000Z

352

High strength uranium-tungsten alloy process  

DOE Patents [OSTI]

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.

Dunn, Paul S. (Santa Fe, NM); Sheinberg, Haskell (Los Alamos, NM); Hogan, Billy M. (Los Alamos, NM); Lewis, Homer D. (Bayfield, CO); Dickinson, James M. (Los Alamos, NM)

1990-01-01T23:59:59.000Z

353

Clean Air Act Requirements: Uranium Mill Tailings  

E-Print Network [OSTI]

EPA'S Clean Air Act Requirements: Uranium Mill Tailings Radon Emissions Rulemaking Reid J. Rosnick requirements for operating uranium mill tailings (Subpart W) Status update on Subpart W activities Outreach/Communications #12;3 EPA Regulatory Requirements for Operating Uranium Mill Tailings (Clean Air Act) · 40 CFR 61

354

URANIUM MILL TAILINGS RADON FLUX CALCULATIONS  

E-Print Network [OSTI]

URANIUM MILL TAILINGS RADON FLUX CALCULATIONS PIĂ?ON RIDGE PROJECT MONTROSE COUNTY, COLORADO Inc. (Golder) was commissioned by EFRC to evaluate the operations of the uranium mill tailings storage in this report were conducted using the WISE Uranium Mill Tailings Radon Flux Calculator, as updated on November

355

Remediation and Recovery of Uranium from Contaminated  

E-Print Network [OSTI]

Remediation and Recovery of Uranium from Contaminated Subsurface Environments with Electrodes K E L that Geobacter species can effectively remove uranium from contaminated groundwater by reducing soluble U was stably precipitated until reoxidized in the presence of oxygen. When an electrode was placed in uranium

Lovley, Derek

356

Uranium Watch REGULATORY CONFUSION: FEDERALAND STATE  

E-Print Network [OSTI]

Uranium Watch Report REGULATORY CONFUSION: FEDERALAND STATE ENFORCEMENT OF 40 C.F.R. PART 61 SUBPART W INTRODUCTION 1. This Uranium Watch Report, Regulatory Confusion: Federal and State Enforcement at the White Mesa Uranium Mill, San Juan County, Utah. 2. The DAQ, a Division of the Utah Department

357

D Riso-R-429 Automated Uranium  

E-Print Network [OSTI]

routinely used analytical techniques for uranium determina- tions in geological samples, fissionCM i D Riso-R-429 Automated Uranium Analysis by Delayed-Neutron Counting H. Kunzendorf, L. Løvborg AUTOMATED URANIUM ANALYSIS BY DELAYED-NEUTRON COUNTING H. Kunzendorf, L. Løvborg and E.M. Christiansen

358

Y-12 Uranium Exposure Study  

SciTech Connect (OSTI)

Following the recent restart of operations at the Y-12 Plant, the Radiological Control Organization (RCO) observed that the enriched uranium exposures appeared to involve insoluble rather than soluble uranium that presumably characterized most earlier Y-12 operations. These observations necessitated changes in the bioassay program, particularly the need for routine fecal sampling. In addition, it was not reasonable to interpret the bioassay data using metabolic parameter values established during earlier Y-12 operations. Thus, the recent urinary and fecal bioassay data were interpreted using the default guidance in Publication 54 of the International Commission on Radiological Protection (ICRP); that is, inhalation of Class Y uranium with an activity median aerodynamic diameter (AMAD) of 1 {micro}m. Faced with apparently new workplace conditions, these actions were appropriate and ensured a cautionary approach to worker protection. As additional bioassay data were accumulated, it became apparent that the data were not consistent with Publication 54. Therefore, this study was undertaken to examine the situation.

Eckerman, K.F.; Kerr, G.D.

1999-08-05T23:59:59.000Z

359

Electric field shielding in dielectric nanosolutions  

E-Print Network [OSTI]

To gain some insight into electrochemical activity of dielectric colloids of technical and biomedical interest we investigate a model of dielectric nanosolution whose micro-constitution is dominated by dipolarions -- positively and negatively charged spherically symmetric nano-structures composed of ionic charge surrounded by cloud of radially polarized dipoles of electrically neutral molecules of solvent. Combing the standard constitutive equations of an isotropic dielectric liquid with Maxwell equation of electrostatics and presuming the Boltzmann shape of the particle density of bound-charge we derive equation for the in-medium electrostatic field. Particular attention is given to numerical analysis of obtained analytic solutions of this equation describing the exterior fields of dipolarions with dipolar atmospheres of solvent molecules endowed with either permanent or field-induced dipole moments radially polarized by central symmetric field of counterions. The presented computations show that the electric field shielding of dipolarions in dielectric nanosolutions is quite different from that of counterionic nano-complexes of Debye-H\\"uckel theory of electrolytes.

Sergey Bastrukov; Pik-Yin Lai; Irina Molodtsova

2014-03-26T23:59:59.000Z

360

Addendum to NuMI shielding assessment  

SciTech Connect (OSTI)

The original safety assessment and the Safety Envelope for the NuMI beam line corresponds to 400 kW of beam power. The Main Injector is currently capable of and approved for producing 500 kW of beam power2. However, operation of the NuMI beam line at 400 kW of power brings up the possibility of an occasional excursion above 400 kW due to better than usual tuning in one of the machines upstream of the NuMI beam line. An excursion above the DOE approved Safety Envelope will constitute a safety violation. The purpose of this addendum is to evaluate the radiological issues and modifications required to operate the NuMI beam line at 500 kW. This upgrade will allow 400 kW operations with a reasonable safety margin. Configuration of the NuMI beam line, boundaries, safety system and the methodologies used for the calculations are as described in the original NuMI SAD. While most of the calculations presented in the original shielding assessment were based on Monte Carlo simulations, which were based on the design geometries, most of the results presented in this addendum are based on the measurements conducted by the AD ES&H radiation safety group.

Vaziri, Kamran; /Fermilab

2007-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Stream depletion by groundwater pumping from leaky Vitaly A. Zlotnik  

E-Print Network [OSTI]

Stream depletion by groundwater pumping from leaky aquifers Vitaly A. Zlotnik Department Maximum Stream Depletion Rate, which is defined as a maximum fraction of the pumping rate supplied focused on hy- draulic connection between a stream and an aquifer for pumping wells in alluvial valleys

Tartakovsky, Daniel M.

362

Pumping induced depletion from two streams Dongmin Sun a  

E-Print Network [OSTI]

Author's personal copy Pumping induced depletion from two streams Dongmin Sun a , Hongbin Zhan b-domain and becomes identical to that of Hunt [Hunt B. Unsteady stream depletion from ground water pumping. Ground of the shortest distance from the pumping well to the other stream over the shortest distance between the two

Zhan, Hongbin

363

Supercontinuum Stimulated Emission Depletion Fluorescence Lifetime Imaging  

SciTech Connect (OSTI)

Supercontinuum (SC) stimulated emission depletion (STED) fluorescence lifetime imaging is demonstrated by using time-correlated single-photon counting (TCSPC) detection. The spatial resolution of the developed STED instrument was measured by imaging monodispersed 40-nm fluorescent beads and then determining their fwhm, and was 36 ± 9 and 40 ± 10 nm in the X and Y coordinates, respectively. The same beads measured by confocal microscopy were 450 ± 50 and 430 ± 30 nm, which is larger than the diffraction limit of light due to underfilling the microscope objective. Underfilling the objective and time gating the signal were necessary to achieve the stated STED spatial resolution. The same fluorescence lifetime (2.0 ± 0.1 ns) was measured for the fluorescent beads by using confocal or STED lifetime imaging. The instrument has been applied to study Alexa Fluor 594-phalloidin labeled F-actin-rich projections with dimensions smaller than the diffraction limit of light in cultured cells. Fluorescence lifetimes of the actin-rich projections range from 2.2 to 2.9 ns as measured by STED lifetime imaging.

Lesoine, Michael; Bose, Sayantan; Petrich, Jacob; Smith, Emily

2012-06-13T23:59:59.000Z

364

Method for the construction of x-ray shielding masks  

SciTech Connect (OSTI)

A method is described for the production of a rigid model of a patient's face onto which lead shielding sheets may be contoured. The model is cast in Lipowitz's metal using a plaster mold.

Canup, D.; Ekstrand, K.E.

1982-03-01T23:59:59.000Z

365

Colorado's Economic Recovery since the Great Recession Professor Martin Shields  

E-Print Network [OSTI]

1 Colorado's Economic Recovery since the Great Recession Professor Martin Shields Regional Economics Institute Colorado State University csurei, economic performance has been mixed. The northern Front Range has fared best

366

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab  

Broader source: Energy.gov [DOE]

American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once used to absorb neutrons produced for...

367

RSMASS: A simple model for estimating reactor and shield masses  

SciTech Connect (OSTI)

A simple mathematical model (RSMASS) has been developed to provide rapid estimates of reactor and shield masses for space-based reactor power systems. Approximations are used rather than correlations or detailed calculations to estimate the reactor fuel mass and the masses of the moderator, structure, reflector, pressure vessel, miscellaneous components, and the reactor shield. The fuel mass is determined either by neutronics limits, thermal/hydraulic limits, or fuel damage limits, whichever yields the largest mass. RSMASS requires the reactor power and energy, 24 reactor parameters, and 20 shield parameters to be specified. This parametric approach should be applicable to a very broad range of reactor types. Reactor and shield masses calculated by RSMASS were found to be in good agreement with the masses obtained from detailed calculations.

Marshall, A.C.; Aragon, J.; Gallup, D.

1987-01-01T23:59:59.000Z

368

Upgrade of the LHC magnet interconnections thermal shielding  

SciTech Connect (OSTI)

The about 1700 interconnections (ICs) between the Large Hadron Collider (LHC) superconducting magnets include thermal shielding at 50-75 K, providing continuity to the thermal shielding of the magnet cryostats to reduce the overall radiation heat loads to the 1.9 K helium bath of the magnets. The IC shield, made of aluminum, is conduction-cooled via a welded bridge to the thermal shield of the adjacent magnets which is actively cooled. TIG welding of these bridges made in the LHC tunnel at installation of the magnets induced a considerable risk of fire hazard due to the proximity of the multi-layer insulation of the magnet shields. A fire incident occurred in one of the machine sectors during machine installation, but fortunately with limited consequences thanks to prompt intervention of the operators. LHC is now undergoing a 2 years technical stop during which all magnet's ICs will have to be opened to consolidate the magnet electrical connections. The IC thermal shields will therefore have to be removed and re-installed after the work is completed. In order to eliminate the risk of fire hazard when re-welding, it has been decided to review the design of the IC shields, by replacing the welded bridges with a mechanical clamping which also preserves its thermal function. An additional advantage of this new solution is the ease in dismantling for maintenance, and eliminating weld-grinding operations at removal needing radioprotection measures because of material activation after long-term operation of the LHC. This paper describes the new design of the IC shields and in particular the theoretical and experimental validation of its thermal performance. Furthermore a status report of the on-going upgrade work in the LHC is given.

Musso, Andrea; Barlow, Graeme; Bastard, Alain; Charrondiere, Maryline; Deferne, Guy; Dib, Gaëlle; Duret, Max; Guinchard, Michael; Prin, Hervé; Craen, Arnaud Vande; Villiger, Gilles [CERN European Organization for Nuclear Research, Meyrin 1211, Geneva 23, CH (Switzerland); Chrul, Anna [The Henryk Niewodniczanski Institute of Nuclear Physics, Polish Academy of Sciences, ul.Radzikowskiego 152, 31-324 Krakow (Poland); Damianoglou, Dimitrios [NTUA National Technical University of Athens, Heeron Polytechniou 9, 15780 Zografou (Greece); Strychalski, Micha? [Wroclaw University of Technology, Faculty of Mechanical and Power Engineering, Wyb. Wyspianskiego 27, Wroclaw, 50-370 (Poland); Wright, Loren [Lancaster University, Bailrigg, Lancaster, LA1 4YW (United Kingdom)

2014-01-29T23:59:59.000Z

369

Electron Backscatter Diffraction (EBSD) Characterization of Uranium and Uranium Alloys  

SciTech Connect (OSTI)

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 of twin boundaries formed during the martensitic transformation. Deformation actually results in a coarsening of the microstructure making EBSD more practical following a limited amount of strain. Figure 2 shows the microstructure resulting from 6% compression. Casting of U-10Mo results in considerable chemical segregation as is apparent in Figure 2a. The segregation subsists through rolling and heat treatment processes as shown in Figure 2b. EBSD was used to study the effects of homogenization time and temperature on chemical heterogeneity. It was found that times and temperatures that result in a chemically homogeneous microstructure also result in a significant increase in grain size. U-0.75Ti forms an acicular martinsite as shown in Figure 4. This microstructure prevails through cycling into the higher temperature solid uranium phases.

McCabe, Rodney J. [Los Alamos National Laboratory; Kelly, Ann Marie [Los Alamos National Laboratory; Clarke, Amy J. [Los Alamos National Laboratory; Field, Robert D. [Los Alamos National Laboratory; Wenk, H. R. [University of California, Berkeley

2012-07-25T23:59:59.000Z

370

Process for alloying uranium and niobium  

DOE Patents [OSTI]

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.

Holcombe, Cressie E. (Farragut, TN); Northcutt, Jr., Walter G. (Oak Ridge, TN); Masters, David R. (Knoxville, TN); Chapman, Lloyd R. (Knoxville, TN)

1991-01-01T23:59:59.000Z

371

Uranium 2014 resources, production and demand  

E-Print Network [OSTI]

Published every other year, Uranium Resources, Production, and Demand, or the "Red Book" as it is commonly known, is jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency. It is the recognised world reference on uranium and is based on official information received from 43 countries. It presents the results of a thorough review of world uranium supplies and demand and provides a statistical profile of the world uranium industry in the areas of exploration, resource estimates, production and reactor-related requirements. It provides substantial new information from all major uranium production centres in Africa, Australia, Central Asia, Eastern Europe and North America. Long-term projections of nuclear generating capacity and reactor-related uranium requirements are provided as well as a discussion of long-term uranium supply and demand issues. This edition focuses on recent price and production increases that could signal major changes in the industry.

Organisation for Economic Cooperation and Development. Paris

2014-01-01T23:59:59.000Z

372

Uranium 2005 resources, production and demand  

E-Print Network [OSTI]

Published every other year, Uranium Resources, Production, and Demand, or the "Red Book" as it is commonly known, is jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency. It is the recognised world reference on uranium and is based on official information received from 43 countries. This 21st edition presents the results of a thorough review of world uranium supplies and demand as of 1st January 2005 and provides a statistical profile of the world uranium industry in the areas of exploration, resource estimates, production and reactor-related requirements. It provides substantial new information from all major uranium production centres in Africa, Australia, Central Asia, Eastern Europe and North America. Projections of nuclear generating capacity and reactor-related uranium requirements through 2025 are provided as well as a discussion of long-term uranium supply and demand issues. This edition focuses on recent price and production increases that could signal major c...

Organisation for Economic Cooperation and Development. Paris

2006-01-01T23:59:59.000Z

373

An analysis of uranium dispersal and health effects using a Gulf War case study.  

SciTech Connect (OSTI)

The study described in this report used mathematical modeling to estimate health risks from exposure to depleted uranium (DU) during the 1991 Gulf War for both U.S. troops and nearby Iraqi civilians. The analysis found that the risks of DU-induced leukemia or birth defects are far too small to result in an observable increase in these health effects among exposed veterans or Iraqi civilians. Only a few veterans in vehicles accidentally struck by U.S. DU munitions are predicted to have inhaled sufficient quantities of DU particulate to incur any significant health risk (i.e., the possibility of temporary kidney damage from the chemical toxicity of uranium and about a 1% chance of fatal lung cancer). The health risk to all downwind civilians is predicted to be extremely small. Recommendations for monitoring are made for certain exposed groups. Although the study found fairly large calculational uncertainties, the models developed and used are generally valid. The analysis was also used to assess potential uranium health hazards for workers in the weapons complex. No illnesses are projected for uranium workers following standard guidelines; nonetheless, some research suggests that more conservative guidelines should be considered.

Marshall, Albert Christian

2005-07-01T23:59:59.000Z

374

A concept of a nonfissile uranium hexafluoride overpack for storage, transport, and processing of corroded cylinders  

SciTech Connect (OSTI)

There is a need to develop a means of safely transporting breached 48-in. cylinders containing depleted uranium hexafluoride (UF{sub 6}) from current storage locations to locations where the contents can be safely removed. There is also a need to provide a method of safely and easily transporting degraded cylinders that no longer meet the US Department of Transportation (DOT) and American National Standards Institute, Inc., (ANSI) requirements for shipments of depleted UF{sub 6}. A study has shown that an overpack can be designed and fabricated to satisfy these needs. The envisioned overpack will handle cylinder models 48G, 48X, and 48Y and will also comply with the ANSI N14.1 and the American Society of Mechanical Engineers (ASME) Sect. 8 requirements.

Pope, R.B.; Cash, J.M. [Oak Ridge National Lab., TN (United States); Singletary, B.H. [Lockheed Martin Energy Systems, Oak Ridge, TN (United States)

1996-06-01T23:59:59.000Z

375

Blue Cross Blue Shield of Massachusetts is an Independent Licensee of the Blue Cross and Blue Shield Association Expanded Coverage for Preventive Care Under  

E-Print Network [OSTI]

Blue Cross Blue Shield of Massachusetts is an Independent Licensee of the Blue Cross and Blue. Effect the New Rules Will Have on Members and Accounts Blue Cross Blue Shield of Massachusetts. Blue Cross Blue Shield of Massachusetts will offer the following services with no member cost share

Aalberts, Daniel P.

376

Global microRNA depletion suppresses tumor angiogenesis  

E-Print Network [OSTI]

MicroRNAs delicately regulate the balance of angiogenesis. Here we show that depletion of all microRNAs suppresses tumor angiogenesis. We generated microRNA-deficient tumors by knocking out Dicer1. These tumors are highly ...

Chen, Sidi

377

The economics of fuel depletion in fast breeder reactor blankets  

E-Print Network [OSTI]

A fast breeder reactor fuel depletion-economics model was developed and applied to a number of 1000 MWe UMBR case studies, involving radial blanket-radial reflector design, radial blanket fuel management, and sensitivity ...

Brewer, Shelby Templeton

1972-01-01T23:59:59.000Z

378

Hyperspectral stimulated emission depletion microscopy and methods of use thereof  

DOE Patents [OSTI]

A hyperspectral stimulated emission depletion ("STED") microscope system for high-resolution imaging of samples labeled with multiple fluorophores (e.g., two to ten fluorophores). The hyperspectral STED microscope includes a light source, optical systems configured for generating an excitation light beam and a depletion light beam, optical systems configured for focusing the excitation and depletion light beams on a sample, and systems for collecting and processing data generated by interaction of the excitation and depletion light beams with the sample. Hyperspectral STED data may be analyzed using multivariate curve resolution analysis techniques to deconvolute emission from the multiple fluorophores. The hyperspectral STED microscope described herein can be used for multi-color, subdiffraction imaging of samples (e.g., materials and biological materials) and for analyzing a tissue by Forster Resonance Energy Transfer ("FRET").

Timlin, Jerilyn A; Aaron, Jesse S

2014-04-01T23:59:59.000Z

379

Reports on investigations of uranium anomalies. National Uranium Resource Evaluation  

SciTech Connect (OSTI)

During the National Uranium Resource Evaluation (NURE) program, conducted for the US Department of Energy (DOE) by Bendix Field Engineering Corporation (BFEC), radiometric and geochemical surveys and geologic investigations detected anomalies indicative of possible uranium enrichment. Data from the Aerial Radiometric and Magnetic Survey (ARMS) and the Hydrogeochemical and Stream-Sediment Reconnaissance (HSSR), both of which were conducted on a national scale, yielded numerous anomalies that may signal areas favorable for the occurrence of uranium deposits. Results from geologic evaluations of individual 1/sup 0/ x 2/sup 0/ quadrangles for the NURE program also yielded anomalies, which could not be adequately checked during scheduled field work. Included in this volume are individual reports of field investigations for the following six areas which were shown on the basis of ARMS, HSSR, and (or) geologic data to be anomalous: (1) Hylas zone and northern Richmond basin, Virginia; (2) Sischu Creek area, Alaska; (3) Goodman-Dunbar area, Wisconsin; (4) McCaslin syncline, Wisconsin; (5) Mt. Withington Cauldron, Socorro County, New Mexico; (6) Lake Tecopa, Inyo County, California. Field checks were conducted in each case to verify an indicated anomalous condition and to determine the nature of materials causing the anomaly. The ultimate objective of work is to determine whether favorable conditions exist for the occurrence of uranium deposits in areas that either had not been previously evaluated or were evaluated before data from recent surveys were available. Most field checks were of short duration (2 to 5 days). The work was done by various investigators using different procedures, which accounts for variations in format in their reports. All papers have been abstracted and indexed.

Goodknight, C.S.; Burger, J.A. (comps.) [comps.

1982-10-01T23:59:59.000Z

380

Global terrestrial uranium supply and its policy implications : a probabilistic projection of future uranium costs  

E-Print Network [OSTI]

An accurate outlook on long-term uranium resources is critical in forecasting uranium costresource relationships, and for energy policy planning as regards the development and deployment of nuclear fuel cycle alternatives. ...

Matthews, Isaac A

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Impact of carbon dioxide sequestration in depleted gas-condensate reservoirs.  

E-Print Network [OSTI]

??Depleted gas-condensate reservoirs are becoming important targets for carbon dioxide sequestration. Since depleted below the dew point, retrograde condensate has been deposited in the pore… (more)

Ramharack, Richard M.

2010-01-01T23:59:59.000Z

382

Treatment of Uranium and Plutonium Solutions Generated in the Atalante Facility, France - 12004  

SciTech Connect (OSTI)

The Atalante complex operated by the French Alternative Energies and Atomic Energy Commission (CEA) at the Rhone Valley Research Center consolidates research programs on actinide chemistry, especially separation chemistry, processing for recycling spent fuel, and fabrication of actinide targets for innovative concepts in future nuclear systems. The design of future systems (Generation IV reactors, material recycling) will increase the uranium and plutonium flows in the facility, making it important to anticipate the stepped-up activity and provide Atalante with equipment dedicated to processing these solutions to obtain a mixed uranium-plutonium oxide that will be stored pending reuse. Ongoing studies for integral recycling of the actinides have highlighted the need for reserving equipment to produce actinides mixed oxide powder and also minor actinides bearing oxide for R and D purpose. To meet this double objective a new shielded line should be built in the facility and should be operational 6 years after go decision. The main functions of the new unit would be to receive, concentrate and store solutions, purify them, ensure group conversion of actinides and conversion of excess uranium. This new unit will be constructed in a completely refurbished building devoted to subcritical and safe geometry of the process equipments. (author)

Lagrave, Herve [French Alternative Energies and Atomic Energy Commission - CEA, Rhone Valley Research Center, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France)

2012-07-01T23:59:59.000Z

383

Uranium 2009 resources, production and demand  

E-Print Network [OSTI]

With several countries currently building nuclear power plants and planning the construction of more to meet long-term increases in electricity demand, uranium resources, production and demand remain topics of notable interest. In response to the projected growth in demand for uranium and declining inventories, the uranium industry – the first critical link in the fuel supply chain for nuclear reactors – is boosting production and developing plans for further increases in the near future. Strong market conditions will, however, be necessary to trigger the investments required to meet projected demand. The "Red Book", jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency, is a recognised world reference on uranium. It is based on information compiled in 40 countries, including those that are major producers and consumers of uranium. This 23rd edition provides a comprehensive review of world uranium supply and demand as of 1 January 2009, as well as data on global ur...

Organisation for Economic Cooperation and Development. Paris

2010-01-01T23:59:59.000Z

384

Uranium Metal Analysis via Selective Dissolution  

SciTech Connect (OSTI)

Uranium metal, which is present in sludge held in the Hanford Site K West Basin, can create hazardous hydrogen atmospheres during sludge handling, immobilization, or subsequent transport and storage operations by its oxidation/corrosion in water. A thorough knowledge of the uranium metal concentration in sludge therefore is essential to successful sludge management and waste process design. The goal of this work was to establish a rapid routine analytical method to determine uranium metal concentrations as low as 0.03 wt% in sludge even in the presence of up to 1000-fold higher total uranium concentrations (i.e., up to 30 wt% and more uranium) for samples to be taken during the upcoming sludge characterization campaign and in future analyses for sludge handling and processing. This report describes the experiments and results obtained in developing the selective dissolution technique to determine uranium metal concentration in K Basin sludge.

Delegard, Calvin H.; Sinkov, Sergey I.; Schmidt, Andrew J.; Chenault, Jeffrey W.

2008-09-10T23:59:59.000Z

385

L'URANIUM ET LES ARMES L'URANIUM APPAUVRI. Pierre Roussel*  

E-Print Network [OSTI]

L'URANIUM ET LES ARMES � L'URANIUM APPAUVRI. Pierre Roussel* Institut de Physique Nucléaire, CNRS massivement dans la guerre du Golfe, des obus anti- chars ont été utilisés, avec des "charges d'uranium, avec une charge de 300 g d'uranium et tiré par des avions, l'autre de 120 mm de diamètre avec une

Boyer, Edmond

386

Shielded fluid stream injector for particle bed reactor  

DOE Patents [OSTI]

A shielded fluid-stream injector assembly is provided for particle bed reactors. The assembly includes a perforated pipe injector disposed across the particle bed region of the reactor and an inverted V-shaped shield placed over the pipe, overlapping it to prevent descending particles from coming into direct contact with the pipe. The pipe and shield are fixedly secured at one end to the reactor wall and slidably secured at the other end to compensate for thermal expansion. An axially extending housing aligned with the pipe and outside the reactor and an in-line reamer are provided for removing deposits from the inside of the pipe. The assembly enables fluid streams to be injected and distributed uniformly into the particle bed with minimized clogging of injector ports. The same design may also be used for extraction of fluid streams from particle bed reactors.

Notestein, John E. (Morgantown, WV)

1993-01-01T23:59:59.000Z

387

Shielding analysis of the NAC-MPC storage system  

SciTech Connect (OSTI)

This paper presents the shielding analyses of the NAC-MPC dry cask storage system. The NAC-MPC dry cask storage system consists of a transportable storage canister, a transfer cask, and a vertical concrete storage cask. The NAC-MPC is designed to accommodate 36 {open_quotes}Yankee Class{close_quotes} fuel assemblies with a maximum burnup of 36,000 MWd/tonne U burnup and 8 yr cooling time. The shielding analysis is performed with the SCALE 4.3 code package which includes SAS2H for source term generation and SAS4A, a modification of SAS4, for shielding evaluations. SAS4 utilizes a one-dimensional XSDRNPM adjoint calculation of the cask to generate biasing parameters for a three-dimensional MORSE-SGC Monte Carlo model of the cask geometry.

Napolitano, D.G.; Romano, N.J. [NAC International, Norcross, GA (United States); Hertel, N.E. [Georgia Institute of Technology, Atlanta, GA (United States)] [and others

1997-12-01T23:59:59.000Z

388

Dry process fluorination of uranium dioxide using ammonium bifluoride  

E-Print Network [OSTI]

An experimental study was conducted to determine the practicality of various unit operations for fluorination of uranium dioxide. The objective was to prepare ammonium uranium fluoride double salts from uranium dioxide and ...

Yeamans, Charles Burnett, 1978-

2003-01-01T23:59:59.000Z

389

SHEEP MOUNTAIN URANIUM PROJECT CROOKS GAP, WYOMING  

E-Print Network [OSTI]

;PROJECT OVERVIEW ·Site Location·Site Location ·Fremont , Wyoming ·Existing Uranium Mine Permit 381C·Existing Uranium Mine Permit 381C ·Historical Operation ·Western Nuclear Crooks Gap Project ·Mined 1956 ­ 1988 and Open Pit Mining ·Current Mine Permit (381C) ·Updating POO, Reclamation Plan & Bond ·Uranium Recovery

390

Review of uranium bioassay techniques  

SciTech Connect (OSTI)

A variety of analytical techniques is available for evaluating uranium in excreta and tissues at levels appropriate for occupational exposure control and evaluation. A few (fluorometry, kinetic phosphorescence analysis, {alpha}-particle spectrometry, neutron irradiation techniques, and inductively-coupled plasma mass spectrometry) have also been demonstrated as capable of determining uranium in these materials at levels comparable to those which occur naturally. Sample preparation requirements and isotopic sensitivities vary widely among these techniques and should be considered carefully when choosing a method. This report discusses analytical techniques used for evaluating uranium in biological matrices (primarily urine) and limits of detection reported in the literature. No cost comparison is attempted, although references are cited which address cost. Techniques discussed include: {alpha}-particle spectrometry; liquid scintillation spectrometry, fluorometry, phosphorometry, neutron activation analysis, fission-track counting, UV-visible absorption spectrophotometry, resonance ionization mass spectrometry, and inductively-coupled plasma mass spectrometry. A summary table of reported limits of detection and of the more important experimental conditions associated with these reported limits is also provided.

Bogard, J.S.

1996-04-01T23:59:59.000Z

391

AXIAL SOURCE PROFILE EFFECT ON WASTE PACKAGE TRANSPORTER SHIELDING  

SciTech Connect (OSTI)

The purpose of this scoping calculation is to support preliminary design of the Waste Package (WP) transporter radiation shield configuration. Spent Nuclear Fuel (SNF) is highly radioactive and site personnel must be protected during the period that the WPs are emplaced. Personnel protection is accomplished via a heavily shielded WP transporter that moves the waste from the surface to the emplacement drift. All previous WP transporter shielding calculations have assumed a Design Basis Fuel (DBF) in which the fuel burnup is uniform (e.g. Ref. 7.3, Ref. 7.4, and Ref. 7.12). In reality, SNF burnup varies significantly from one end of the fuel assembly to the other. Since source strengths are dependent upon fuel burnup, a model which varies the fuel burnup along the assembly axis will produce a more accurate depiction of the radiation field surrounding the WP transporter. The objective of this calculation is to determine the need for using the actual axial profile, as opposed to the uniform burnup assumption, in the WP transporter shield design. The scope of the calculation is as follows: (1) Determine the impact of axial source term variation on WP transporter contact dose rates. (2) Determine appropriate shielding modifications to account for expected dose rate peaking effects. Consistent with the previous subsurface shielding analyses, this calculation considers the bounding 21 Pressurized Water Reactor (PWR) WP only. The calculation will need to be revised and extended to Boiling Water Reactor (BWR) SNF upon selection of the WP design for the License Application (LA) and availability of the source terms from the WP Operations Group.

A. Nielsen

1999-06-30T23:59:59.000Z

392

Uranium Tris-aryloxide Derivatives Supported by Triazacyclononane: Engendering a Reactive Uranium(III)  

E-Print Network [OSTI]

, we are currently investigating the coordina- tion chemistry of uranium metal centers with classicalUranium Tris-aryloxide Derivatives Supported by Triazacyclononane: Engendering a Reactive Uranium, and Karsten Meyer* Contribution from the Department of Chemistry and Biochemistry, UniVersity of California

Meyer, Karsten

393

Statistical data of the uranium industry  

SciTech Connect (OSTI)

Statistical Data of the Uranium Industry 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 1981. 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 (GJAO) of the US Department of Energy. The production, reserves, and drilling information is reported in a manner which avoids disclosure of proprietary information.

none,

1982-01-01T23:59:59.000Z

394

Adsorptive Stripping Voltammetric Measurements of Trace Uranium...  

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

Measurements of Trace Uranium at the Bismuth Film Electrode. Abstract: Bismuth-coated carbon-fiber electrodes have been successfully applied for adsorptive-stripping...

395

Biogeochemical Processes In Ethanol Stimulated Uranium Contaminated...  

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

A laboratory incubation experiment was conducted with uranium contaminated subsurface sediment to assess the geochemical and microbial community response to ethanol amendment. A...

396

Colorimetric detection of uranium in water  

DOE Patents [OSTI]

Disclosed are methods, materials and systems that can be used to determine qualitatively or quantitatively the level of uranium contamination in water samples. Beneficially, disclosed systems are relatively simple and cost-effective. For example, disclosed systems can be utilized by consumers having little or no training in chemical analysis techniques. Methods generally include a concentration step and a complexation step. Uranium concentration can be carried out according to an extraction chromatographic process and complexation can chemically bind uranium with a detectable substance such that the formed substance is visually detectable. Methods can detect uranium contamination down to levels even below the MCL as established by the EPA.

DeVol, Timothy A. (Clemson, SC); Hixon, Amy E. (Piedmont, SC); DiPrete, David P. (Evans, GA)

2012-03-13T23:59:59.000Z

397

Uranium Weapons Components Successfully Dismantled | National...  

National Nuclear Security Administration (NNSA)

Successfully Dismantled March 20, 2007 Uranium Weapons Components Successfully Dismantled Oak Ridge, TN Continuing its efforts to reduce the size of the U.S. nuclear weapons...

398

High strength and density tungsten-uranium alloys  

DOE Patents [OSTI]

Alloys of tungsten and uranium and a method for making the alloys. The amount of tungsten present in the alloys is from about 55 vol % to about 85 vol %. A porous preform is made by sintering consolidated tungsten powder. The preform is impregnated with molten uranium such that (1) uranium fills the pores of the preform to form uranium in a tungsten matrix or (2) uranium dissolves portions of the preform to form a continuous uranium phase containing tungsten particles.

Sheinberg, Haskell (Los Alamos, NM)

1993-01-01T23:59:59.000Z

399

Distribution of uranium-bearing phases in soils from Fernald  

SciTech Connect (OSTI)

Electron beam techniques have been used to characterize uranium-contaminated soils and the Fernald Site, Ohio. Uranium particulates have been deposited on the soil through chemical spills and from the operation of an incinerator plant on the site. The major uranium phases have been identified by electron microscopy as uraninite, autunite, and uranium phosphite [U(PO{sub 3}){sub 4}]. Some of the uranium has undergone weathering resulting in the redistribution of uranium within the soil.

Buck, E.C.; Brown, N.R.; Dietz, N.L.

1993-12-31T23:59:59.000Z

400

Co-Designing Sustainable Communities: The Identification and Incorporation of Social Performance Metrics in Native American Sustainable Housing and Renewable Energy System Design  

E-Print Network [OSTI]

of uranium and depleted uranium exposure on reproduction andTeratogenicity of depleted uranium aerosols: A review from

Shelby, Ryan

2013-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

President Truman Increases Production of Uranium and Plutonium...  

National Nuclear Security Administration (NNSA)

Increases Production of Uranium and Plutonium October 09, 1950 President Truman Increases Production of Uranium and Plutonium Washington, DC President Truman approves a 1.4...

402

Atomistic Simulations of Uranium Incorporation into Iron (Hydr...  

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

of Uranium Incorporation into Iron (Hydr)Oxides. Atomistic Simulations of Uranium Incorporation into Iron (Hydr)Oxides. Abstract: Atomistic simulations were carried out to...

403

Toxic Substances Control Act Uranium Enrichment Federal Facility...  

Office of Environmental Management (EM)

Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic...

404

Geochemical Controls on Contaminant Uranium in Vadose Hanford...  

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

Controls on Contaminant Uranium in Vadose Hanford Formation Sediments at the 200 Area and 300 Area, Hanford Site, Geochemical Controls on Contaminant Uranium in Vadose Hanford...

405

Microbial Reduction of Uranium under Iron- and Sulfate-reducing...  

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

Uranium under Iron- and Sulfate-reducing Conditions: Effect of Amended Goethite on Microbial Community Microbial Reduction of Uranium under Iron- and Sulfate-reducing Conditions:...

406

Uncertainty analysis of multi-rate kinetics of uranium desorption...  

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

Uncertainty analysis of multi-rate kinetics of uranium desorption from sediments. Uncertainty analysis of multi-rate kinetics of uranium desorption from sediments. Abstract: A...

407

Legacy Management Work Progresses on Defense-Related Uranium...  

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

Most recently, LM visited 84 defense-related legacy uranium mine sites located within 11 uranium mining districts in 6 western states. At these sites, photographs and global...

408

Highly Enriched Uranium Materials Facility, Major Design Changes...  

Energy Savers [EERE]

Highly Enriched Uranium Materials Facility, Major Design Changes Late...Lessons Learned Report, NNSA, Dec 2010 Highly Enriched Uranium Materials Facility, Major Design Changes...

409

Record of Decision for the Uranium Leasing Program Programmatic...  

Energy Savers [EERE]

Record of Decision for the Uranium Leasing Program Programmatic Environmental Impact Statement Record of Decision for the Uranium Leasing Program Programmatic Environmental Impact...

410

DOE Extends Public Comment Period for the Draft Uranium Leasing...  

Office of Environmental Management (EM)

Extends Public Comment Period for the Draft Uranium Leasing Program Programmatic Environmental Impact Statement DOE Extends Public Comment Period for the Draft Uranium Leasing...

411

Sequestering Uranium from Seawater: Binding Strength and Modes...  

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

Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl Complexes with Glutarimidedioxime Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl...

412

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

SciTech Connect (OSTI)

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

NONE

1996-07-01T23:59:59.000Z

413

Analytical solution for Joule-Thomson cooling during CO2 geo-sequestration in depleted oil and gas reservoirs  

E-Print Network [OSTI]

sequestration in depleted oil and gas reservoirs Simon A.1. Introduction Depleted oil and gas reservoirs (DOGRs)

Mathias, S.A.

2010-01-01T23:59:59.000Z

414

aperture shield materials: Topics by E-print Network  

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

aperture shield materials First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Light-weight Flexible...

415

A Note on Hamilton Cycles in Kneser Graphs Ian Shields  

E-Print Network [OSTI]

A Note on Hamilton Cycles in Kneser Graphs Ian Shields IBM P.O. Box 12195 Research Triangle Park) have Hamilton cycles when n #20; 27. A similar result is shown for bipartite Kneser graphs. 1 for Hamilton cycles in Kneser graphs, K(n; k), and bipartite Kneser graphs, H(n; k). With the exception

Savage, Carla D.

416

RSMASS-D: Reactor and shield mass minimization models  

SciTech Connect (OSTI)

Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid metal cooled reactors (LMR's), in-core thermionic reactors (TI's) and out-of-core thermionic reactors (OTR's). The approach was based on much of the methodology developed for the RSMASS model (Marshall 1986). The models use a combination of simple equations derived from reactor physics and other fundamental considerations along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within an allowed range to achieve a parameter choice which yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage and thermal limitations are accounted for in the computations. Thermionic requirements are also accounted for in the thermionic reactor models. All major reactor component masses are estimated as well as instrumentation and control (I C), boom and safety system mass. A new shield model was developed and incorporated into all three models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from detailed calculations within 16 percent for all three models.

Marshall, A.C. (Department 0410 Sandia National Laboratories (USA) NE-52 The Department of Energy, Germantown Building, Washington, D.C. 20545 (USA)); Gallup, D.R. (Division 6472 Sandia National Laboratories, P.O. Box 5800, Albuquerque, New Mexico 87185 (USA))

1991-01-01T23:59:59.000Z

417

February, 2010 Fire Analysis of the Shielded Container  

E-Print Network [OSTI]

Pilot Plant Carlsbad, New Mexico Preparer: Independent Technical Review: Ray Sprankle RSL Safety;FIRE ANALYSIS OF THE SHIELDED CONTAINER FOR THE WIPP-032, REV. 0 WASTE ISOLATION PILOT PLANT, CARLSBAD, CARLSBAD, NEW MEXICO PAGE 3 OF 28 Executive Summary Transuranic waste is currently shipped to the WIPP from

418

RZ calculations for self shielded multigroup cross sections  

SciTech Connect (OSTI)

A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l'Energie Atomique CEA, Direction de l'Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)

2006-07-01T23:59:59.000Z

419

PERGAMON Carbon 39 (2001) 279285 Electromagnetic interference shielding effectiveness of carbon  

E-Print Network [OSTI]

PERGAMON Carbon 39 (2001) 279­285 Review Electromagnetic interference shielding effectiveness materials for electromagnetic interference (EMI) shielding are reviewed. They include composite materials-structural and structural composites, colloi- dal graphite, as well as EMI gasket materials. Electromagnetic interference

Chung, Deborah D.L.

420

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask  

SciTech Connect (OSTI)

The VSC-17 Spent Nuclear Fuel Storage Cask was surveyed for degradation of the concrete shield by radiation measurement, temperature measurement, and ultrasonic testing. No general loss of shielding function was identified.

Koji Shirai

2006-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

Accelerator shield design of KIPT neutron source facility  

SciTech Connect (OSTI)

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary at less than 0.5-mrem/hr. The shield configuration and parameters of the accelerator building have been determined and are presented in this paper. (authors)

Zhong, Z.; Gohar, Y. [Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States)

2013-07-01T23:59:59.000Z

422

Bioremediation of uranium contaminated soils and wastes  

SciTech Connect (OSTI)

Contamination of soils, water, and sediments by radionuclides and toxic metals from uranium mill tailings, nuclear fuel manufacturing and nuclear weapons production is a major concern. Studies of the mechanisms of biotransformation of uranium and toxic metals under various microbial process conditions has resulted in the development of two treatment processes: (1) stabilization of uranium and toxic metals with reduction in waste volume and (2) removal and recovery of uranium and toxic metals from wastes and contaminated soils. Stabilization of uranium and toxic metals in wastes is accomplished by exploiting the unique metabolic capabilities of the anaerobic bacterium, Clostridium sp. The radionuclides and toxic metals are solubilized by the bacteria directly by enzymatic reductive dissolution, or indirectly due to the production of organic acid metabolites. The radionuclides and toxic metals released into solution are immobilized by enzymatic reductive precipitation, biosorption and redistribution with stable mineral phases in the waste. Non-hazardous bulk components of the waste volume. In the second process uranium and toxic metals are removed from wastes or contaminated soils by extracting with the complexing agent citric acid. The citric-acid extract is subjected to biodegradation to recover the toxic metals, followed by photochemical degradation of the uranium citrate complex which is recalcitrant to biodegradation. The toxic metals and uranium are recovered in separate fractions for recycling or for disposal. The use of combined chemical and microbiological treatment process is more efficient than present methods and should result in considerable savings in clean-up and disposal costs.

Francis, A.J.

1998-12-31T23:59:59.000Z

423

Uranium Management - Preservation of a National Asset  

SciTech Connect (OSTI)

The Uranium Management Group (UMG) was established at the Department of Energy's (DOE's) Oak Ridge Operations in 1999 as a mechanism to expedite the de-inventory of surplus uranium from the Fernald Environmental Management Project site. This successful initial venture has broadened into providing uranium material de-inventory and consolidation support to the Hanford site as well as retrieving uranium materials that the Department had previously provided to universities under the loan/lease program. As of December 31, 2001, {approx} 4,300 metric tons of uranium (MTU) have been consolidated into a more cost effective interim storage location at the Portsmouth site near Piketon, OH. The UMG continues to uphold its corporate support mission by promoting the Nuclear Materials Stewardship Initiative (NMSI) and the twenty-five (25) action items of the Integrated Nuclear Materials Management Plan (1). Before additional consolidation efforts may commence to remove excess inventory from Environmental Management closure sites and universities, a Programmatic Environmental Assessment (PEA) must be completed. Two (2) noteworthy efforts currently being pursued involve the investigation of re-use opportunities for surplus uranium materials and the recovery of usable uranium from the shutdown Portsmouth cascade. In summary, the UMG is available as a DOE complex-wide technical resource to promote the responsible management of surplus uranium.

Jackson, J. D.; Stroud, J. C.

2002-02-27T23:59:59.000Z

424

Method for fabricating laminated uranium composites  

DOE Patents [OSTI]

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

Chapman, L.R.

1983-08-03T23:59:59.000Z

425

Scrap uranium recycling via electron beam melting  

SciTech Connect (OSTI)

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.

McKoon, R.

1993-11-01T23:59:59.000Z

426

National Uranium Resource Evaluation, Tonopah quadrangle, Nevada  

SciTech Connect (OSTI)

The Tonopah Quadrangle, Nevada, was evaluated using National Uranium Resource Evaluation criteria to identify and delineate areas favorable for uranium deposits. Investigations included reconnaissance and detailed surface geologic and radiometric studies, geochemical sampling and evaluation, analysis and ground-truth followup of aerial radiometric and hydrogeochemical and stream-sediment reconnaissance data, and subsurface data evaluation. The results of these investigations indicate environments favorable for hydroallogenic uranium deposits in Miocene lacustrine sediments of the Big Smoky Valley west of Tonopah. The northern portion of the Toquima granitic pluton is favorable for authigenic uranium deposits. Environments considered unfavorable for uranium deposits include Quaternary sediments; intermediate and mafic volcanic and metavolcanic rocks; Mesozoic, Paleozoic, and Precambrian sedimentary and metasedimentary rocks; those plutonic rocks not included within favorable areas; and those felsic volcanic rocks not within the Northumberland and Mount Jefferson calderas.

Hurley, B W; Parker, D P

1982-04-01T23:59:59.000Z

427

Uranium in prehistoric Indian pottery  

E-Print Network [OSTI]

present in the sample, and the cross l section of the process (the measure of the probability of a neutron interacting with an uranium atom), In general, a daughter product 235 of U fission is analyzed on a detector which counts either gamma rays... for quantitative analysis of various elements on archaeological artifacts, Manganese has been determined in Mesoamerican pot sherds (Bennyhoff and Heizer 1965). A Pu-Be radioisotope neutron source with a flux of 4 x 10 4 -2 -1 neutrons cm sec was used...

Filberth, Ernest William

2012-06-07T23:59:59.000Z

428

2013 Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"Click worksheet9,1,50022,3,,,,6,1,,781Title: Telephone:short version)ec 1827190List of Tables3 Uranium

429

2013 Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"Click worksheet9,1,50022,3,,,,6,1,,781Title: Telephone:short version)ec 1827190List of Tables3 Uranium11

430

2013 Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"Click worksheet9,1,50022,3,,,,6,1,,781Title: Telephone:short version)ec 1827190List of6,2009Uranium

431

2013 Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"Click worksheet9,1,50022,3,,,,6,1,,781Title: Telephone:short version)ec 1827190List of6,2009UraniumNext

432

2013 Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"Click worksheet9,1,50022,3,,,,6,1,,781Title: Telephone:short version)ec 1827190List of6,2009UraniumNext

433

U.S.Uranium Reserves  

Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelinesProved ReservesFeet)per Thousand28 198 18BiomassThree-Dimensional SeismicUranium

434

2013 Uranium Marketing Annual Report  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to:SeadovCooperativeA2. World liquids consumption by region,Purchases2 U.S.Feed6a. Uranium

435

2013 Uranium Marketing Annual Report  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to:SeadovCooperativeA2. World liquids consumption by region,Purchases2 U.S.Feed6a.4. Uranium

436

A low-noise ferrite magnetic shield T. W. Kornack,a  

E-Print Network [OSTI]

A low-noise ferrite magnetic shield T. W. Kornack,a S. J. Smullin, S.-K. Lee, and M. V. Romalis April 2007; published online 29 May 2007 Ferrite materials provide magnetic shielding performance by thermal Johnson currents due to their high electrical resistivity. Measurements inside a ferrite shield

Romalis, Mike

437

Electromagnetic interference shielding reaching 70 dB in steel fiber cement  

E-Print Network [OSTI]

Electromagnetic interference shielding reaching 70 dB in steel fiber cement Sihai Wen, D.D.L. Chung; Silica fume; Shielding 1. Introduction Electromagnetic interference (EMI) shielding [1­4] is in critical, NY 14260-4400, USA Received 9 January 2002; accepted 14 August 2003 Abstract An electromagnetic

Chung, Deborah D.L.

438

Blue Shield ensures uninterrupted access to quality medical care after Palm Drive Hospital ceases operations  

E-Print Network [OSTI]

Blue Shield ensures uninterrupted access to quality medical care after Palm Drive Hospital ceases and inpatient services due to the hospital's bankruptcy filing. This closure affects Blue Shield HMO plan members who have been utilizing Palm Drive Hospital services. Please be advised that Blue Shield

Ravikumar, B.

439

Company Name: Blue Cross Blue Shield of MA Web Site: www.bluecrossma.com  

E-Print Network [OSTI]

Company Name: Blue Cross Blue Shield of MA Web Site: www.bluecrossma.com Industry: Healthcare Brief Company Overview: Headquartered in Boston, Blue Cross Blue Shield of Massachusetts provides comprehensive-level position: Please visit www.bluecrossma.com/careers. With almost 3 million members, Blue Cross Blue Shield

New Hampshire, University of

440

Colorado Supplement to the Summary of Benefits and Coverage Form Anthem Blue Cross and Blue Shield  

E-Print Network [OSTI]

Supplement to the Summary of Benefits and Coverage Form Anthem Blue Cross and Blue Shield CHEIBA Prime PPO Screenings, Prostate Cancer Screenings, and Colorectal Cancer Screenings. Anthem Blue Cross and Blue Shield Colorado, Inc. Independent licensees of the Blue Cross and Blue Shield Association. ®ANTHEM is a registered

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


441

Colorado Supplement to the Summary of Benefits and Coverage Form Anthem Blue Cross and Blue Shield  

E-Print Network [OSTI]

Supplement to the Summary of Benefits and Coverage Form Anthem Blue Cross and Blue Shield CHEIBA Custom Plus Screenings, Prostate Cancer Screenings, and Colorectal Cancer Screenings. Anthem Blue Cross and Blue Shield Colorado, Inc. Independent licensees of the Blue Cross and Blue Shield Association. ®ANTHEM is a registered

442

Flood and Shield Basalts from Ethiopia: Magmas from the African Superswell  

E-Print Network [OSTI]

Flood and Shield Basalts from Ethiopia: Magmas from the African Superswell BRUNO KIEFFER1, ETHIOPIA 4 DEEPARTEMENT DES SCIENCES DE LA TERRE ET DE L'ENVIRONNEMENT, UNIVERSITEE LIBRE DE BRUXELLES 50 the shield volcanoes. KEY WORDS: Ethiopia; flood basalts; shield volcanism; superswell INTRODUCTION According

Demouchy, Sylvie

443

Draft Environmental Impact Statement for Construction and Operation of a Depleted Uranium Hexafluoride Conversion Facility at Portsmouth, Ohio, Site  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergy CooperationRequirements Matrix U.S.7685 Vol. 76, No. 29DoingSRS-WD-2010-001 Revision 04:2360

444

Removal of uranium from uranium-contaminated soils -- Phase 1: Bench-scale testing. Uranium in Soils Integrated Demonstration  

SciTech Connect (OSTI)

To address the management of uranium-contaminated soils at Fernald and other DOE sites, the DOE Office of Technology Development formed the Uranium in Soils Integrated Demonstration (USID) program. The USID has five major tasks. These include the development and demonstration of technologies that are able to (1) characterize the uranium in soil, (2) decontaminate or remove uranium from the soil, (3) treat the soil and dispose of any waste, (4) establish performance assessments, and (5) meet necessary state and federal regulations. This report deals with soil decontamination or removal of uranium from contaminated soils. The report was compiled by the USID task group that addresses soil decontamination; includes data from projects under the management of four DOE facilities [Argonne National Laboratory (ANL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), and the Savannah River Plant (SRP)]; and consists of four separate reports written by staff at these facilities. The fundamental goal of the soil decontamination task group has been the selective extraction/leaching or removal of uranium from soil faster, cheaper, and safer than current conventional technologies. The objective is to selectively remove uranium from soil without seriously degrading the soil`s physicochemical characteristics or generating waste forms that are difficult to manage and/or dispose of. Emphasis in research was placed more strongly on chemical extraction techniques than physical extraction techniques.

Francis, C. W.

1993-09-01T23:59:59.000Z

445

Recovery of uranium by using new microorganisms isolated from North American uranium deposits  

SciTech Connect (OSTI)

Some attempts were made to remove uranium that may be present in refining effluents, mine tailings by using new microorganisms isolated from uranium deposits and peculiar natural environments. To screen microorganisms isolated from uranium deposits and peculiar natural environments in North America and Japan for maximal accumulation of uranium, hundreds of microorganisms were examined. Some microorganisms can accumulate about 500 mg (4.2 mEq) of uranium per gram of Microbial cells within 1 h. The uranium accumulation capacity of the cells exceeds that of commercially available chelating agents (2-3 mEq/g adsorbent). We attempted to recover uranium from uranium refining waste water by using new microorganisms. As a result, these microbial cells can recover trace amounts of uranium from uranium waste water with high efficiency. These strains also have a high accumulating ability for thorium. Thus, these new microorganisms can be used as an adsorbing agent for the removal of nuclear elements may be present in metallurgical effluents, mine tailings and other waste sources.

Sakaguchi, T.; Nakajima, A.; Tsuruta, T. [Miyazaki Medical College (Japan)

1995-12-31T23:59:59.000Z

446

Uranium Cluster Chemistry DOI: 10.1002/anie.200906605  

E-Print Network [OSTI]

Uranium Cluster Chemistry DOI: 10.1002/anie.200906605 Tetranuclear Uranium Clusters by Reductive in the coordination chemistry and small-molecule reactivity of uranium. Among the intriguing reactivity patterns of tetravalent uranium with 3,5-dimethylpyrazolate (Me2PzĂ? ) led to forma- tion of an unprecedented homoleptic

447

Learning about ozone depletion Paul J. Crutzen & Michael Oppenheimer  

E-Print Network [OSTI]

Learning about ozone depletion Paul J. Crutzen & Michael Oppenheimer Received: 12 January 2007 Mainz, Germany M. Oppenheimer (*) Department of Geosciences, Princeton University, Princeton, NJ 08544, USA e-mail: omichael@princeton.edu M. Oppenheimer Woodrow Wilson School of Public and International

Oppenheimer, Michael

448

Lithium depletion and the rotational history of exoplanet host stars  

E-Print Network [OSTI]

Israelian et al. (2004) reported that exoplanet host stars are lithium depleted compared to solar-type stars without detected massive planets, a result recently confirmed by Gonzalez (2008). We investigate whether enhanced lithium depletion in exoplanet host stars may result from their rotational history. We have developed rotational evolution models for slow and fast solar-type rotators from the pre-main sequence (PMS) to the age of the Sun and compare them to the distribution of rotational periods observed for solar-type stars between 1 Myr and 5 Gyr. We show that slow rotators develop a high degree of differential rotation between the radiative core and the convective envelope, while fast rotators evolve with little core-envelope decoupling. We suggest that strong differential rotation at the base of the convective envelope is responsible for enhanced lithium depletion in slow rotators. We conclude that lithium-depleted exoplanet host stars were slow rotators on the zero-age main sequence (ZAMS) and argue that slow rotation results from a long lasting star-disk interaction during the PMS. Altogether, this suggests that long-lived disks (> 5 Myr) may be a necessary condition for massive planet formation/migration.

Jerome Bouvier

2008-09-03T23:59:59.000Z

449

Hypolimnetic Oxygen Depletion in Eutrophic Lakes Beat Muller,*,  

E-Print Network [OSTI]

Hypolimnetic Oxygen Depletion in Eutrophic Lakes Beat Muller,*, Lee D. Bryant,, Andreas Matzinger obtained from 11 eutrophic lakes and suggests a model describing the consumption of dissolved oxygen (O2) in the hypolimnia of eutrophic lakes as a result of only two fundamental processes: O2 is consumed (i) by settled

Wehrli, Bernhard

450

The Variation of Magnesium Depletion with Line of Sight Conditions  

E-Print Network [OSTI]

In this paper we report on the gas-phase abundance of singly-ionized magnesium (Mg II) in 44 lines of sight, using data from the Hubble Space Telescope (HST). We measure Mg II column densities by analyzing medium- and high-resolution archival STIS spectra of the 1240 A doublet of Mg II. We find that Mg II depletion is correlated with many line of sight parameters (e.g. F(H_2), E_(B-V), E_(B-V)/r, A_V, and A_V/r) in addition to the well-known correlation with . These parameters should be more directly related to dust content and thus have more physical significance with regard to the depletion of elements such as magnesium. We examine the significance of these additional correlations as compared to the known correlation between Mg II depletion and . While none of the correlations are better predictors of Mg II depletion than , some are statistically significant even assuming fixed . We discuss the ranges over which these correlations are valid, their strength at fixed , and physical interpretations.

Adam G. Jensen; Theodore P. Snow

2007-10-04T23:59:59.000Z

451

Defending Resource Depletion Attacks on Implantable Medical Devices  

E-Print Network [OSTI]

and storage. In this research, we identify a new kind of attacks on IMDs - Resource Depletion (RD) attacks information. IMD attacks may also be launched by insurance companies. IMD readers may be installed near, and storage. An IMD is implanted in patient's body and expected to run for several years. Typical IMDs

Wu, Jie

452

Field corrosion testing and performance of cable shielding materials in soils  

SciTech Connect (OSTI)

This article discusses the importance of corrosion resistance in cable-shielding materials, describes the mechanisms of shielding corrosion that occur in buried telephone cable, and evaluates the results of the six-year REA Horry Cooperative buried telephone cable corrosion test. In this study, both active and static cables were included. Withdrawals were made over a six-year period. These cables were evaluated for cable-shielding corrosion. Special attention was paid to the comparative behavior of active and static cables. Results indicate that steel shieldings are most susceptible to the effects of alternating current (AC) in active cables. Results of a wide range of shieldings are presented and evaluated.

Haynes, G.; Baboian, R. (Texas Instruments Inc., Electrochemical and Corrosion Lab., 34 Forest St., Mail Station 10-13, Attleboro, MA (US))

1989-09-01T23:59:59.000Z

453

Technical Basis for Assessing Uranium Bioremediation Performance  

SciTech Connect (OSTI)

In situ bioremediation of uranium holds significant promise for effective stabilization of U(VI) from groundwater at reduced cost compared to conventional pump and treat. This promise is unlikely to be realized unless researchers and practitioners successfully predict and demonstrate the long-term effectiveness of uranium bioremediation protocols. Field research to date has focused on both proof of principle and a mechanistic level of understanding. Current practice typically involves an engineering approach using proprietary amendments that focuses mainly on monitoring U(VI) concentration for a limited time period. Given the complexity of uranium biogeochemistry and uranium secondary minerals, and the lack of documented case studies, a systematic monitoring approach using multiple performance indicators is needed. This document provides an overview of uranium bioremediation, summarizes design considerations, and identifies and prioritizes field performance indicators for the application of uranium bioremediation. The performance indicators provided as part of this document are based on current biogeochemical understanding of uranium and will enable practitioners to monitor the performance of their system and make a strong case to clients, regulators, and the public that the future performance of the system can be assured and changes in performance addressed as needed. The performance indicators established by this document and the information gained by using these indicators do add to the cost of uranium bioremediation. However, they are vital to the long-term success of the application of uranium bioremediation and provide a significant assurance that regulatory goals will be met. The document also emphasizes the need for systematic development of key information from bench scale tests and pilot scales tests prior to full-scale implementation.

PE Long; SB Yabusaki; PD Meyer; CJ Murray; AL N’Guessan

2008-04-01T23:59:59.000Z

454

Prediction of effective atomic number (Z) for laminated shielding material  

E-Print Network [OSTI]

(&?/P)=3. 103E-2 (P/P)=5. 951E-2 (P, ?/P)=2. 603 E-2 (&/P)=6. 803E-2 P', ?/P)=3. 654E-2 3. 0 (&/P)=3. 968E-2 (&, ?/P)=2. 281E-2 (4/P)=3. 616E-2 (P, ?/P)=2, 042E 2 (P/P)=4. 200E-2 (&??/P)=2. 322E-2 25 35 30 25 20 N 15 10 ~ffe w ~2fe w ~3... fe w ~4fe w ~5fe w 0 0 4 mfp 10 Fig. 5a. Z?fr for two-layered shields (Zf&Z2) as a function of mfp 90 80 70 4D ~2w pb 3D ~3w pb 20 ~4w pb 10 ~6w pb 0 0 2 4 6 8 1D m fp Fig. 5. b. Z, fr for two-layered shields (Zf&Z2) as a function...

Sarder, Md. Maksudur Rahaman

1999-01-01T23:59:59.000Z

455

Superconducting shielded core reactor with reduced AC losses  

DOE Patents [OSTI]

A superconducting shielded core reactor (SSCR) operates as a passive device for limiting excessive AC current in a circuit operating at a high power level under a fault condition such as shorting. The SSCR includes a ferromagnetic core which may be either closed or open (with an air gap) and extends into and through a superconducting tube or superconducting rings arranged in a stacked array. First and second series connected copper coils each disposed about a portion of the iron core are connected to the circuit to be protected and are respectively wound inside and outside of the superconducting tube or rings. A large impedance is inserted into the circuit by the core when the shielding capability of the superconducting arrangement is exceeded by the applied magnetic field generated by the two coils under a fault condition to limit the AC current in the circuit. The proposed SSCR also affords reduced AC loss compared to conventional SSCRs under continuous normal operation.

Cha, Yung S.; Hull, John R.

2006-04-04T23:59:59.000Z

456

CORROSION OF LEAD SHIELDING IN NUCLEAR MATERIALS PACKAGES  

SciTech Connect (OSTI)

Inspection of United States-Department of Energy (US-DOE) model 9975 nuclear materials shipping package revealed corrosion of the lead shielding that was induced by off-gas constituents from organic components in the package. Experiments were performed to determine the corrosion rate of lead when exposed to off-gas or degradation products of these organic materials. The results showed that the room temperature vulcanizing (RTV) sealant was the most corrosive organic species used in the construction of the packaging, followed by polyvinyl acetate (PVAc) glue. Fiberboard material, also used in the construction of the packaging induced corrosion to a much lesser extent than the PVAc glue and RTV sealant, and only in the presence of condensed water. The results indicated faster corrosion at temperatures higher than ambient and with condensed water. In light of these corrosion mechanisms, the lead shielding was sheathed in a stainless steel liner to mitigate against corrosion.

Subramanian, K; Kerry Dunn, K; Joseph Murphy, J

2008-07-18T23:59:59.000Z

457

Electrochemistry, Spectroscopy, and Reactivity of Uranium Complexes Supported by Ferrocene Diamide Ligands  

E-Print Network [OSTI]

J. L. , Pentavalent Uranium Chemistry-Synthetic Pursuit of afor Trivalent Uranium Chemistry. Inorg. Chem. 1989, 28, (and High-Valent Uranium Chemistry. Organometallics 2011,

Duhovic, Selma

2012-01-01T23:59:59.000Z

458

Recent International R&D Activities in the Extraction of Uranium from Seawater  

E-Print Network [OSTI]

Uranium and Rare Earth Elements Using Biomass of Algae, Bioinorganic ChemistryRecovery of uranium from sea water. Chemistry & Industry (uranium recovery from seawater. Industrial & Engineering Chemistry

Rao, Linfeng

2011-01-01T23:59:59.000Z

459

Electrolytic process for preparing uranium metal  

DOE Patents [OSTI]

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.

Haas, Paul A. (Knoxville, TN)

1990-01-01T23:59:59.000Z

460

Attenuation of high-energy x rays by iron shielding  

SciTech Connect (OSTI)

Monte Carlo calculations are presented on electron-accelerator x-ray spectra for actual target thicknesses and electron energies of 4-50 MeV. Effective attenuation coefficients have been obtained as well as build-up factors for collimated beams andiron shielding of thickness form 1 to 80 cm. The radiation contrast has been determined as a function of thickness for this energy range.

Bespalov, V.I.; Chakhlov, V.L.; Shtein, M.M.

1988-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "depleted uranium shielded" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


461

ACCURATE TEMPERATURE MEASUREMENTS IN A NATURALLY-ASPIRATED RADIATION SHIELD  

SciTech Connect (OSTI)

Experiments and calculations were conducted with a 0.13 mm fine wire thermocouple within a naturally-aspirated Gill radiation shield to assess and improve the accuracy of air temperature measurements without the use of mechanical aspiration, wind speed or radiation measurements. It was found that this thermocouple measured the air temperature with root-mean-square errors of 0.35 K within the Gill shield without correction. A linear temperature correction was evaluated based on the difference between the interior plate and thermocouple temperatures. This correction was found to be relatively insensitive to shield design and yielded an error of 0.16 K for combined day and night observations. The correction was reliable in the daytime when the wind speed usually exceeds 1 m s{sup -1} but occasionally performed poorly at night during very light winds. Inspection of the standard deviation in the thermocouple wire temperature identified these periods but did not unambiguously locate the most serious events. However, estimates of sensor accuracy during these periods is complicated by the much larger sampling volume of the mechanically-aspirated sensor compared with the naturally-aspirated sensor and the presence of significant near surface temperature gradients. The root-mean-square errors therefore are upper limits to the aspiration error since they include intrinsic sensor differences and intermittent volume sampling differences.

Kurzeja, R.

2009-09-09T23:59:59.000Z

462

Crystal Chemistry of Early Actinides (Thorium, Uranium, and Neptunium) and Uranium Mesoporous Materials.  

E-Print Network [OSTI]

??Despite their considerable global importance, the structural chemistry of actinides remains understudied. Thorium and uranium fuel cycles are used in commercial nuclear reactors in India… (more)

Sigmon, Ginger E.

2010-01-01T23:59:59.000Z

463

Prokaryotic microorganisms in uranium mining waste piles and their interactions with uranium and other heavy metals.  

E-Print Network [OSTI]

??The influence of uranyl and sodium nitrate under aerobic and anaerobic conditions on the microbial community structure of a soil sample from the uranium mining… (more)

Geißler, Andrea

2007-01-01T23:59:59.000Z

464

CRDIAC: Coupled Reactor Depletion Instrument with Automated Control  

SciTech Connect (OSTI)

When modeling the behavior of a nuclear reactor over time, it is important to understand how the isotopes in the reactor will change, or transmute, over that time. This is especially important in the reactor fuel itself. Many nuclear physics modeling codes model how particles interact in the system, but do not model this over time. Thus, another code is used in conjunction with the nuclear physics code to accomplish this. In our code, Monte Carlo N-Particle (MCNP) codes and the Multi Reactor Transmutation Analysis Utility (MRTAU) were chosen as the codes to use. In this way, MCNP would produce the reaction rates in the different isotopes present and MRTAU would use cross sections generated from these reaction rates to determine how the mass of each isotope is lost or gained. Between these two codes, the information must be altered and edited for use. For this, a Python 2.7 script was developed to aid the user in getting the information in the correct forms. This newly developed methodology was called the Coupled Reactor Depletion Instrument with Automated Controls (CRDIAC). As is the case in any newly developed methodology for modeling of physical phenomena, CRDIAC needed to be verified against similar methodology and validated against data taken from an experiment, in our case AFIP-3. AFIP-3 was a reduced enrichment plate type fuel tested in the ATR. We verified our methodology against the MCNP Coupled with ORIGEN2 (MCWO) method and validated our work against the Post Irradiation Examination (PIE) data. When compared to MCWO, the difference in concentration of U-235 throughout Cycle 144A was about 1%. When compared to the PIE data, the average bias for end of life U-235 concentration was about 2%. These results from CRDIAC therefore agree with the MCWO and PIE data, validating and verifying CRDIAC. CRDIAC provides an alternative to using ORIGEN-based methodology, which is useful because CRDIAC's depletion code, MRTAU, uses every available isotope in its depletion, unlike ORIGEN, which only depletes the isotopes specified by the user. This means that depletions done by MRTAU more accurately reflect reality. MRTAU also allows the user to build new isotope data sets, which means any isotope with nuclear data could be depleted, something that would help predict the outcomes of nuclear reaction testing in materials other than fuel, like beryllium or gold.

Steven K. Logan

2012-08-01T23:59:59.000Z

465

In situ remediation of uranium contaminated groundwater  

SciTech Connect (OSTI)

In an effort to develop cost-efficient techniques for remediating uranium contaminated groundwater at DOE Uranium Mill Tailing Remedial Action (UMTRA) sites nationwide, Sandia National Laboratories (SNL) deployed a pilot scale research project at an UMTRA site in Durango, CO. Implementation included design, construction, and subsequent monitoring of an in situ passive reactive barrier to remove Uranium from the tailings pile effluent. A reactive subsurface barrier is produced by emplacing a reactant material (in this experiment - various forms of metallic iron) in the flow path of the contaminated groundwater. Conceptually the iron media reduces and/or adsorbs uranium in situ to acceptable regulatory levels. In addition, other metals such as Se, Mo, and As have been removed by the reductive/adsorptive process. The primary objective of the experiment was to eliminate the need for surface treatment of tailing pile effluent. Experimental design, and laboratory and field preliminary results are discussed with regard to other potential contaminated groundwater treatment applications.

Dwyer, B.P.; Marozas, D.C. [Sandia National Labs., Albuquerque, NM (United States)

1997-12-31T23:59:59.000Z

466

In situ remediation of uranium contaminated groundwater  

SciTech Connect (OSTI)

In an effort to develop cost-efficient techniques for remediating uranium contaminated groundwater at DOE Uranium Mill Tailing Remedial Action (UMTRA) sites nationwide, Sandia National Laboratories (SNL) deployed a pilot scale research project at an UMTRA site in Durango, CO. Implementation included design, construction, and subsequent monitoring of an in situ passive reactive barrier to remove Uranium from the tailings pile effluent. A reactive subsurface barrier is produced by emplacing a reactant material (in this experiment various forms of metallic iron) in the flow path of the contaminated groundwater. Conceptually the iron media reduces and/or adsorbs uranium in situ to acceptable regulatory levels. In addition, other metals such as Se, Mo, and As have been removed by the reductive/adsorptive process. The primary objective of the experiment was to eliminate the need for surface treatment of tailing pile effluent. Experimental design, and laboratory and field results are discussed with regard to other potential contaminated groundwater treatment applications.

Dwyer, B.P.; Marozas, D.C.

1997-02-01T23:59:59.000Z

467

Process for reducing beta activity in uranium  

DOE Patents [OSTI]

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

Briggs, Gifford G. (Cincinnatti, OH); Kato, Takeo R. (Cincinnatti, OH); Schonegg, Edward (Cleves, OH)

1986-01-01T23:59:59.000Z

468

Method of recovering uranium from aqueous solution  

SciTech Connect (OSTI)

Anion exchange resin derived from insoluble crosslinked polymers of vinyl benzyl chloride which are prepared by polymerizing vinyl benzyl chloride and a crosslinking monomer are particularly suitable in the treatment of uranium bearing leach liquors.

Albright, R.L.

1980-01-22T23:59:59.000Z

469

Innovative design of uranium startup fast reactors  

E-Print Network [OSTI]

Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

Fei, Tingzhou

2012-01-01T23:59:59.000Z

470

Process for reducing beta activity in uranium  

DOE Patents [OSTI]

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

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

1985-04-11T23:59:59.000Z

471

Blue Cross and Blue Shield of Florida is now Florida Blue State Employees' PPO Plan health insurance provider Blue Cross and Blue Shield of  

E-Print Network [OSTI]

Blue Cross and Blue Shield of Florida is now Florida Blue State Employees' PPO Plan health insurance provider Blue Cross and Blue Shield of Florida has recently changed its name to Florida Blue Resources Benefits Office at (850) 6444015, or insben@admin.fsu.edu. RELATED LINKS ­ Florida Blue

Ronquist, Fredrik

472

Benchmarking FENDL libraries through analysis of bulk shielding experiments on large SS316 assemblies for verification of ITER shielding characteristics  

SciTech Connect (OSTI)

FENDL-1 data base has been developed recently for use in ITER/EDA phase and other fusion-related design activities. It is now undergoing extensive testing and benchmarking using experimental data of differential and integral measured parameters obtained from fusion-oriented experiments. As part of co-operation between UCLA (U.S.) with JAERI (Japan) on executing the required neutronics R&D tasks for ITER shield design, two bulk shielding experiments on large SS316 assemblies were selected for benchmarking FENDL/MG-1 multigroup data base and FENDL/MC-1 continous energy data base. The analyses with the multigroup data (performed with S8, P5, DORT calculations with shielded and unshielded data) also included library derived from ENDF/B-VI data base for comparison purposes. The MCNP Monte Carlo code was used by JAERI with the FENDL/MC-1 data. The results of this benchmarking is reported in this paper along with the observed deficiencies and discrepancies. 20 refs., 27 figs., 1 tab.

Youssef, M.Z.; Kumar, A.; Abdou, M.A. [Univ. of California, Los Angeles, CA (United States); Konno, Chikara; Maekawa, Fujio; Wada, Masayuki; Oyama, Yukio; Maekawa, Hiroshi; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Ibaraki (Japan)

1996-12-31T23:59:59.000Z

473

WAPDEG Analysis of Waste Package and Drip shield Degradation  

SciTech Connect (OSTI)

As directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), an analysis of the degradation of the engineered barrier system (EBS) drip shields and waste packages at the Yucca Mountain repository is developed. The purpose of this activity is to provide the TSPA with inputs and methodologies used to evaluate waste package and drip shield degradation as a function of exposure time under exposure conditions anticipated in the repository. This analysis provides information useful to satisfy ''Yucca Mountain Review Plan, Final Report'' (NRC 2003 [DIRS 163274]) requirements. Several features, events, and processes (FEPs) are also discussed (Section 6.2, Table 15). The previous revision of this report was prepared as a model report in accordance with AP-SIII.10Q, Models. Due to changes in the role of this report since the site recommendation, it no longer contains model development. This revision is prepared as a scientific analysis in accordance with AP-SIII.9Q, ''Scientific Analyses'' and uses models previously validated in (1) ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]); (2) ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' (BSC 2004 [DIRS 169984]); and (3) ''General Corrosion and Localized Corrosion of Drip Shield'' (BSC 2004 [DIRS 169845]). The integrated waste package degradation (IWPD) analysis presented in this report treats several implementation-related issues, such as defining the number and size of patches per waste package that undergo stress corrosion cracking; recasting the weld flaw analysis in a form as implemented in the Closure Weld Defects (CWD) software; and, general corrosion rate manipulations (e.g., change of scale in Section 6.3.4). The weld flaw portion of this report takes input from an engineering calculation (BSC 2004 [DIRS 170024]) and uses standard mathematical methods to enable easier implementation. The IWPD analysis also provides guidance on implementation of early failures (importance sampling and multinomial distribution usage). These manipulations are evident from standard scientific practices, approaches, or methods and do not require changes to the previously validated models. The IWPD analysis itself (Section 6.4), not the resultant curves from executing the IWPD analysis presented in Section 6.5 (which are for illustrative purposes), is used directly in total system performance assessment (TSPA). The IWPD analysis simulates general corrosion and stress corrosion cracking of the waste package outer barrier and general corrosion of the drip shield. The effects of igneous and seismic events and localized corrosion on drip shield and waste package performance are not evaluated in this report. The outputs of this report are inputs and methodologies used by TSPA to evaluate waste package and drip shield degradation as a function of exposure time under exposure conditions anticipated in the repository. The analyses presented in this report are for the current repository design (BSC 2004 [DIRS 168489]).

K. Mon

2004-09-29T23:59:59.000Z

474

BIOREMEDIATION OF URANIUM CONTAMINATED SOILS AND WASTES.  

SciTech Connect (OSTI)

Contamination of soils, water, and sediments by radionuclides and toxic metals from uranium mill tailings, nuclear fuel manufacturing and nuclear weapons production is a major concern. Studies of the mechanisms of biotransformation of uranium and toxic metals under various microbial process conditions has resulted in the development of two treatment processes: (i) stabilization of uranium and toxic metals with reduction in waste volume and (ii) removal and recovery of uranium and toxic metals from wastes and contaminated soils. Stabilization of uranium and toxic metals in wastes is accomplished by exploiting the unique metabolic capabilities of the anaerobic bacterium, Clostridium sp. The radionuclides and toxic metals are solubilized by the bacteria directly by enzymatic reductive dissolution, or indirectly due to the production of organic acid metabolites. The radionuclides and toxic metals released into solution are immobilized by enzymatic reductive precipitation, biosorption and redistribution with stable mineral phases in the waste. Non-hazardous bulk components of the waste such as Ca, Fe, K, Mg and Na released into solution are removed, thus reducing the waste volume. In the second process uranium and toxic metals are removed from wastes or contaminated soils by extracting with the complexing agent citric acid. The citric-acid extract is subjected to biodegradation to recover the toxic metals, followed by photochemical degradation of the uranium citrate complex which is recalcitrant to biodegradation. The toxic metals and uranium are recovered in separate fractions for recycling or for disposal. The use of combined chemical and microbiological treatment process is more efficient than present methods and should result in considerable savings in clean-up and disposal costs.

FRANCIS,A.J.

1998-09-17T23:59:59.000Z

475

Material property correlations for uranium mononitride  

E-Print Network [OSTI]

MATERIAL PROPERTY CORRELATIONS FOR URANIUM MONONITRIDE A Thesis by STEVEN LOWE HAYES Submitted to the Office of Graduate Studies of Texas ARM University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE August... 1989 Major Subject: Nuclear Engineering MATERIAL PROPERTY CORRELATIONS FOR URANIUM MONONITRIDE A Thesis by STEVEN LOWE HAYES Approved as to style and content by: K. L. Peddicord (Chair of Committee) R. R. Hart (Member) C. P. Burger (Member...

Hayes, Steven Lowe

2012-06-07T23:59:59.000Z

476

Petrochemical and Mineralogical Constraints on the Source and Processes of Uranium Mineralisation in the Granitoids of Zing-Monkin Area, Adamawa Massif, NE Nigeria  

SciTech Connect (OSTI)

Zing-Monkin area, located in the northern part of Adamawa Massif, is underlain by extensive exposures of moderately radioactive granodiorites, anatectic migmatites, equigranular granites, porphyritic granites and highly radioactive fine-grained granites with minor pegmatites. Selected major and trace element petrochemical investigations of the rocks show that a progression from granodiorite through migmatite to granites is characterised by depletion of MgO, CaO, Fe{sub 2}O{sub 3,} Sr, Ba, and Zr, and enrichment of SiO{sub 2} and Rb. This trend is associated with uranium enrichment and shows a chemical gradation from the more primitive granodiorite to the more evolved granites. Electron microprobe analysis shows that the uranium is content in uranothorite and in accessories, such as monazite, titanite, apatite, epidote and zircon. Based on petrochemical and mineralogical data, the more differentiated granitoids (e.g., fine-grained granite) bordering the Benue Trough are the immediate source of the uranium prospect in Bima Sandstone within the Trough. Uranium was derived from the granitoids by weathering and erosion. Transportation and subsequent interaction with organic matter within the Bima Sandstone led to precipitation of insoluble secondary uranium minerals in the Benue Trough.

Haruna, I. V., E-mail: vela_hi@yahoo.co.uk [Federal University of Technology, Geology Department (Nigeria); Orazulike, D. M. [Abubakar Tafawa Balewa University, Geology Programme (Nigeria); Ofulume, A. B. [Federal University of Technology, Geosciences Department (Nigeria); Mamman, Y. D. [Federal University of Technology, Geology Department (Nigeria)