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1

Nuclear Fuel Facts: Uranium | Department of Energy  

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

Uranium Management and Uranium Management and Policy » Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Uranium is a silvery-white metallic chemical element in the periodic table, with atomic number 92. It is assigned the chemical symbol U. A uranium atom has 92 protons and 92 electrons, of which 6 are valence electrons. Uranium has the highest atomic weight (19 kg m) of all naturally occurring elements. Uranium occurs naturally in low concentrations in soil, rock and water, and is commercially extracted from uranium-bearing minerals such as uraninite. Uranium ore can be mined from open pits or underground excavations. The ore can then be crushed and treated at a mill to separate the valuable uranium from the ore. Uranium may also be dissolved directly from the ore deposits

2

Sizing particles of natural uranium and nuclear fuels using poly-allyl-diglycol carbonate autoradiography  

Science Journals Connector (OSTI)

......particles of natural uranium and nuclear fuels...low enriched, depleted and natural uranium and also aged...committed doses and cancer risks(4...Bristol, UK, sized uranium fragments found...nuclear fuels of depleted uranium (depUO2......

G. Hegyi; R. B. Richardson

2008-07-01T23:59:59.000Z

3

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

4

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

5

Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle  

Science Journals Connector (OSTI)

The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and ...

Frank A. Settle

2009-03-01T23:59:59.000Z

6

THERMODYNAMIC MODEL FOR URANIUM DIOXIDE BASED NUCLEAR FUEL  

SciTech Connect (OSTI)

Many projects involving nuclear fuel rest on a quantitative understanding of the co-existing phases at various stages of burnup. Since the many fission products have considerably different abilities to chemically associate with oxygen, and the oxygen-to-metal molar ratio is slowly changing, the chemical potential of oxygen is a function of burnup. Concurrently, well-recognized small fractions of new phases such as inert gas, noble metals, zirconates, etc. also develop. To further complicate matters, the dominant UO2 fuel phase may be non-stoichiometric and most of the minor phases themselves have a variable composition dependent on temperature and possible contact with the coolant in the event of a sheathing breach. A thermodynamic fuel model to predict the phases in partially burned CANDU (CANada Deuterium Uranium) nuclear fuel containing many major fission products has been under development. The building blocks of the model are the standard Gibbs energies of formation of the many possible compounds expressed as a function of temperature. To these data are added mixing terms associated with the appearance of the component species in particular phases. In operational terms, the treatment rests on the ability to minimize the Gibbs energy in a multicomponent system, in our case using the algorithms developed by Eriksson. The model is capable of handling non-stoichiometry in the UO2 fluorite phase, dilute solution behaviour of significant solute oxides, noble metal inclusions, a second metal solid solution U(Pd-Rh-Ru)3, zirconate, molybdate, and uranate solutions as well as other minor solid phases, and volatile gaseous species.

Thompson, Dr. William T. [Royal Military College of Canada; Lewis, Dr. Brian J [Royal Military College of Canada; Corcoran, E. C. [Royal Military College of Canada; Kaye, Dr. Matthew H. [Royal Military College of Canada; White, S. J. [Royal Military College of Canada; Akbari, F. [Atomic Energy of Canada Limited, Chalk River Laboratories; Higgs, Jamie D. [Atomic Energy of Canada Limited, Point Lepreau; Thompson, D. M. [Praxair Inc.; Besmann, Theodore M [ORNL; Vogel, S. C. [Los Alamos National Laboratory (LANL)

2007-01-01T23:59:59.000Z

7

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

SciTech Connect (OSTI)

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

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

2011-11-01T23:59:59.000Z

8

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

9

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

SciTech Connect (OSTI)

A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface.

Travelli, A.

1988-01-19T23:59:59.000Z

10

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

SciTech Connect (OSTI)

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

none,

2013-07-01T23:59:59.000Z

11

RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA  

SciTech Connect (OSTI)

In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

2009-07-01T23:59:59.000Z

12

Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel  

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

Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel Vitaly T. Gotovchikov a , Victor A. Seredenko a , Valentin V. Shatalov a , Vladimir N. Kaplenkov a , Alexander S. Shulgin a , Vladimir K. Saranchin a , Michail A. Borik a∗ , Charles W. Forsberg b , All-Russian Research Institute of Chemical Technology (ARRICT) 33, Kashirskoe ave., Moscow, Russia, 115409, E-mail: chem.conv@ru.net Oak Ridge National Laboratory (ORNL) Bethel Wall Road, P.O. Box 2008, MS-6165, Oak Ridge, TN, USA, 37831 Abstract - Induction cold crucible melters (ICCM) have the potential to be a very-low-cost high-throughput method for the production of DUO 2 for SNF casks. The proposed work would develop these melters for this specific application. If a

13

Fusion solution to dispose of spent nuclear fuel, transuranic elements, and highly enriched uranium  

Science Journals Connector (OSTI)

The disposal of the nuclear spent fuel, the transuranic elements, and the highly enriched uranium represents a major problem under investigation by the international scientific community to identify the most promising solutions. The investigation of this paper focused on achieving the top rated solution for the problem, the elimination goal, which requires complete elimination for the transuranic elements or the highly enriched uranium, and the long-lived fission products. To achieve this goal, fusion blankets with liquid carrier, molten salts or liquid metal eutectics, for the transuranic elements and the uranium isotopes are utilized. The generated energy from the fusion blankets is used to provide revenue for the system. The long-lived fission products are fabricated into fission product targets for transmutation utilizing the neutron leakage from the fusion blankets. This paper investigated the fusion blanket designs for small fusion devices and the system requirements for such application. The results show that 334 MW of fusion power from D–T plasma for 30 years with an availability factor of 0.75 can dispose of the 70,000 tons of the U.S. inventory of spent nuclear fuel generated up to the year 2015. In addition, this fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future.

Yousry Gohar

2001-01-01T23:59:59.000Z

14

Sizing particles of natural uranium and nuclear fuels using poly-allyl-diglycol carbonate autoradiography  

Science Journals Connector (OSTI)

......University Health Center, Montreal...Biology and Health Physics Branch...enriched, depleted and natural uranium and also aged...fiberglass filter. Health Phys (2000...of natural uranium and nuclear...enriched, depleted and natural......

G. Hegyi; R. B. Richardson

2008-07-01T23:59:59.000Z

15

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

SciTech Connect (OSTI)

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

Collins, Robert T [ORNL] [ORNL; Collins, Jack Lee [ORNL] [ORNL; Hunt, Rodney Dale [ORNL] [ORNL; Ladd-Lively, Jennifer L [ORNL] [ORNL; Patton, Kaara K [ORNL] [ORNL; Hickman, Robert [NASA Marshall Space Flight Center, Huntsville, AL] [NASA Marshall Space Flight Center, Huntsville, AL

2014-01-01T23:59:59.000Z

16

Atomic Diffusion in the Uranium-50wt% Zirconium Nuclear Fuel System  

E-Print Network [OSTI]

Atomic diffusion phenomena were examined in a metal-alloy nuclear fuel system composed of ?-phase U-50wt%Zr fuel in contact with either Zr-10wt%Gd or Zr-10wt%Er. Each alloy was fabricated from elemental feed material via melt-casting, and diffusion...

Eichel, Daniel

2013-06-17T23:59:59.000Z

17

EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany  

Broader source: Energy.gov [DOE]

This environmental assessment (EA) will evaluate the potential environmental impacts of a DOE proposal to accept used nuclear fuel from the Federal Republic of Germany at DOE’s Savannah River Site (SRS) for processing and disposition. This used nuclear fuel is composed of kernels containing thorium and U.S.-origin highly enriched uranium (HEU) embedded in small graphite spheres that were irradiated in nuclear reactors used for research and development purposes.

18

Nuclear Fuels | Department of Energy  

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

Nuclear Fuels Nuclear Fuels Nuclear Fuels A reactor's ability to produce power efficiently is significantly affected by the composition and configuration of its fuel system. A nuclear fuel assembly consists of hundreds of thousands of uranium pellets, stacked and encapsulated within tubes called fuel rods or fuel pins which are then bundled together in various geometric arrangements. There are many design considerations for the material composition and geometric configuration of the various components comprising a nuclear fuel system. Future designs for the fuel and the assembly or packaging of fuel will contribute to cleaner, cheaper and safer nuclear energy. Today's process for developing and testing new fuel systems is resource and time intensive. The process to manufacture the fuel, build an assembly,

19

Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors  

SciTech Connect (OSTI)

The ANSI/ANS 8.12 standard was first approved in July 1978. At that time, this edition was applicable to operations with plutonium-uranium oxide (MOX) fuel mixtures outside reactors and was limited to subcritical limits for homogeneous systems. The next major revision, ANSI/ANS-8.12-1987, included the addition of subcritical limits for heterogeneous systems. The standard was subsequently reaffirmed in February 1993. During late 1990s, substantial work was done by the ANS 8.12 Standard Working Group to re-examine the technical data presented in the standard using the latest codes and cross section sets. Calculations performed showed good agreement with the values published in the standard. This effort resulted in the reaffirmation of the standard in March 2002. The standard is currently in a maintenance mode. After 2002, activities included discussions to determine the future direction of the standard and to follow the MOX standard development by the International Standard Organization (ISO). In 2007, the Working Group decided to revise the standard to extend the areas of applicability by providing a wider range of subcritical data. The intent is to cover a wider domain of MOX fuel fabrication and operations. It was also decided to follow the ISO MOX standard specifications (related to MOX density and isotopics) and develop a new set of subcritical limits for homogeneous systems. This has resulted in the submittal (and subsequent approval) of the project initiation notification system form (PINS) in 2007.

Biswas, D; Mennerdahl, D

2008-06-23T23:59:59.000Z

20

Nuclear Fuels  

Science Journals Connector (OSTI)

The core of a nuclear reactor is composed of a controlled critical configuration of a fissile material, which in strict a sense is the fuel. This fissile material is contained in a matrix, normally a ceramic c...

Rudy J. M. Konings; Thierry Wiss…

2011-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Nuclear & Uranium - U.S. Energy Information Administration (EIA)  

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

Nuclear & Uranium Nuclear & Uranium Glossary › FAQS › Overview Data Summary Uranium & Nuclear Fuel Nuclear Power Plants Radioactive Waste International All Nuclear Data Reports Analysis & Projections Most Requested Nuclear Plants and Reactors Projections Uranium All Reports EIA's latest Short-Term Energy Outlook for electricity › chart showing U.S. electricity generation by fuel, all sectors Source: U.S. Energy Information Administration, Short-Term Energy Outlook, released monthly. Quarterly uranium production data › image chart of Quarterly uranium production as described in linked report Source: U.S. Energy Information Administration, Domestic Uranium Production Report - Quarterly, 3rd Quarter 2013, October 31, 2013. Uprates can increase U.S. nuclear capacity substantially without building

22

Standard test method for analysis of isotopic composition of uranium in nuclear-grade fuel material by quadrupole inductively coupled plasma-mass spectrometry  

E-Print Network [OSTI]

1.1 This test method is applicable to the determination of the isotopic composition of uranium (U) in nuclear-grade fuel material. The following isotopic weight percentages are determined using a quadrupole inductively coupled plasma-mass spectrometer (Q-ICP-MS): 233U, 234U, 235U, 236U, and 238U. The analysis can be performed on various material matrices after acid dissolution and sample dilution into water or dilute nitric (HNO3) acid. These materials include: fuel product, uranium oxide, uranium oxide alloys, uranyl nitrate (UNH) crystals, and solutions. The sample preparation discussed in this test method focuses on fuel product material but may be used for uranium oxide or a uranium oxide alloy. Other preparation techniques may be used and some references are given. Purification of the uranium by anion-exchange extraction is not required for this test method, as it is required by other test methods such as radiochemistry and thermal ionization mass spectroscopy (TIMS). This test method is also described i...

American Society for Testing and Materials. Philadelphia

2000-01-01T23:59:59.000Z

23

Thorium–based fuel cycles : saving uranium in a 200 MWth pebble bed high temperature reactor / S.K. Gintner.  

E-Print Network [OSTI]

??The predominant nuclear fuel used globally at present is uranium which is a finite resource. Thorium has been identified as an alternative nuclear fuel source… (more)

Gintner, Stephan Konrad

2010-01-01T23:59:59.000Z

24

Separation of uranium from technetium in recovery of spent nuclear fuel  

DOE Patents [OSTI]

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

Friedman, H.A.

1984-06-13T23:59:59.000Z

25

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… (more)

Garnetti, David J.

2010-01-01T23:59:59.000Z

26

Nuclear Fuel Cycle | Department of Energy  

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

Cycle Cycle Nuclear Fuel Cycle This is an illustration of a nuclear fuel cycle that shows the required steps to process natural uranium from ore for preparation for fuel to be loaded in nuclear reactors. This is an illustration of a nuclear fuel cycle that shows the required steps to process natural uranium from ore for preparation for fuel to be loaded in nuclear reactors. The mission of NE-54 is primarily focused on activities related to the front end of the nuclear fuel cycle which includes mining, milling, conversion, and enrichment. Uranium Mining Both "conventional" open pit, underground mining, and in situ techniques are used to recover uranium ore. In general, open pit mining is used where deposits are close to the surface and underground mining is used

27

Compositions and methods for treating nuclear fuel  

DOE Patents [OSTI]

Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

2013-08-13T23:59:59.000Z

28

Compositions and methods for treating nuclear fuel  

DOE Patents [OSTI]

Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

2014-01-28T23:59:59.000Z

29

Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center  

SciTech Connect (OSTI)

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

Myers, Astasia [Stanford University, Stanford, CA 94305, USA and MonAme Scientific Research Center, Ulaanbaatar (Mongolia)

2011-06-28T23:59:59.000Z

30

Winter Heating Fuels - Energy Information Administration  

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

stocks, imports and exports. Renewable & Alternative Fuels Includes hydropower, solar, wind, geothermal, biomass and ethanol. Nuclear & Uranium Uranium fuel, nuclear...

31

US-Russian collaboration for enhancing nuclear materials protection, control, and accounting at the Elektrostal uranium fuel-fabrication plant  

SciTech Connect (OSTI)

In September 1993, an implementing agreement was signed that authorized collaborative projects to enhance Russian national materials control and accounting, physical protection, and regulatory activities, with US assistance funded by the Nunn-Lugar Act. At the first US-Russian technical working group meeting in Moscow in February 1994, it was decided to identify a model facility where materials protection, control, and accounting (MPC and A) and regulatory projects could be carried out using proven technologies and approaches. The low-enriched uranium (LEU or RBMK and VVER) fuel-fabrication process at Elektrostal was selected, and collaborative work began in June 1994. Based on many factors, including initial successes at Elektrostal, the Russians expanded the cooperation by proposing five additional sites for MPC and A development: the Elektrostal medium-enriched uranium (MEU or BN) fuel-fabrication process and additional facilities at Podolsk, Dmitrovgrad, Obninsk, and Mayak. Since that time, multilaboratory teams have been formed to develop and implement MPC and A upgrades at the additional sites, and much new work is underway. This paper summarizes the current status of MPC and A enhancement projects in the LEU fuel-fabrication process and discusses the status of work that addresses similar enhancements in the MEU (BN) fuel processes at Elektrostal, under the recently expanded US-Russian MPC and A cooperation.

Smith, H. [Los Alamos National Lab., NM (United States); Allentuck, J. [Brookhaven National Lab., Upton, NY (United States); Barham, M. [Oak Ridge National Lab., TN (United States); Bishop, M. [Sandia National Labs., Albuquerque, NM (United States); Wentz, D. [Lawrence Livermore National Lab., CA (United States); Steele, B.; Bricker, K. [Pacific Northwest National Lab., Richland, WA (United States); Cherry, R. [USDOE, Washington, DC (United States); Snegosky, T. [Dept. of Defense, Washington, DC (United States). Defense Nuclear Agency

1996-09-01T23:59:59.000Z

32

Nuclear Fuel Cycle | Department of Energy  

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

Fuel Cycle Fuel Cycle Nuclear Fuel Cycle GC-52 provides legal advice to DOE regarding research and development of nuclear fuel and waste management technologies that meet the nation's energy supply, environmental, and energy security needs. GC-52 also advises DOE on issues involving support for international fuel cycle initiatives aimed at advancing a common vision of the necessity of the expansion of nuclear energy for peaceful purposes worldwide in a safe and secure manner. In addition, GC-52 provides legal advice to DOE regarding the management and disposition of excess uranium in DOE's uranium stockpile. GC-52 attorneys participate in meetings of DOE's Uranium Inventory Management Coordinating Committee and provide advice on compliance with statutory requirements for the sale or transfer of uranium.

33

Nuclear Reactor Materials and Fuels  

Science Journals Connector (OSTI)

Nuclear reactor materials and fuels can be classified into six categories: Nuclear fuel materials Nuclear clad materials Nuclear coolant materials Nuclear poison materials Nuclear moderator materials

Dr. James S. Tulenko

2012-01-01T23:59:59.000Z

34

Advanced nuclear fuel  

SciTech Connect (OSTI)

Kurt Terrani uses his expertise in materials science to develop safer fuel for nuclear power plants.

Terrani, Kurt

2014-07-14T23:59:59.000Z

35

Advanced nuclear fuel  

ScienceCinema (OSTI)

Kurt Terrani uses his expertise in materials science to develop safer fuel for nuclear power plants.

Terrani, Kurt

2014-07-15T23:59:59.000Z

36

Neptunium - Uranium - Plutonium Co-Extraction in TBP-based Solvent Extraction Processes for Spent Nuclear Fuel Recycling  

SciTech Connect (OSTI)

The US, through the Global Nuclear Energy Partnership, is currently engaged in efforts aimed at closing the nuclear fuel cycle. Neptunium behavior is important to understand for transuranic recycling because of its complex oxidation chemistry. The Pacific Northwest National Laboratory is investigating neptunium oxidation chemistry in the context of the PUREX process. Neptunium extraction in the PUREX process relies on maintaining either IV or V oxidation states. Qualitative conversion of neptunium(V) to neptunium(VI) was achieved within 5 hours in 6 M nitric acid at 95 deg. C. However, the VI state was not maintained during a batch contact test simulating the PUREX process and neptunium reduced to the V state, rendering it inextractable. Vanadium(V) was found to be effective in maintaining neptunium(VI) by adding it to a simulated irradiated nuclear fuel feed in 6 M nitric acid and to the scrub acid in the batch contact simulation of the PUREX process. Computer simulations of the PUREX process with a typical irradiated nuclear fuel in 6 M nitric acid as feed indicated little impact of the higher acid concentration on the behavior of fission products of moderate extractability. We plan to perform countercurrent tests of this modified PUREX process in the near future. (authors)

Arm, S.T.; Abrefah, J.; Lumetta, G.J.; Sinkov, S.I. [Battelle PNWD, Pacific Northwest National Laboratory, 902 Battelle Boulevard, PO Box 999, Richland, Washington, 99352 (United States)

2007-07-01T23:59:59.000Z

37

Plutonium partitioning in uranium and plutonium co-recovery system for fast reactor fuel recycling with enhanced nuclear proliferation resistance  

SciTech Connect (OSTI)

For enhancement of nuclear proliferation resistance, a 'co-processing' method for U and Pu co-recovery was studied. Two concepts, no U scrubbing and no Pu reduction partitioning, were employed to formulate two types of flow sheets by using a calculation code. Their process performance was demonstrated using radioactive solutions derived from an irradiated fast reactor fuel. These experimental results indicated that U and Pu were co-recovered in the U/Pu product, and the Pu content in the U/Pu product increased approximately 2.3 times regardless of using reductant. The proposed no U scrubbing and no Pu reductant flow sheet is applicable to fast reactor fuel reprocessing and enhances its resistance to nuclear proliferation. (authors)

Nakahara, Masaumi; Koma, Yoshikazu; Nakajima, Yasuo [Japan Atomic Energy Agency: 4-33 Muramatsu, Naka-gun, Tokai-mura, Ibaraki 319-1194 (Japan)

2013-07-01T23:59:59.000Z

38

Nuclear Fuel Cycle Integrated System Analysis  

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

Fuel Cycle Integrated System Analysis Fuel Cycle Integrated System Analysis Abdellatif M. Yacout Argonne National Laboratory Nuclear Engineering Division The nuclear fuel cycle is a complex system with multiple components and activities that are combined to provide nuclear energy to a variety of end users. The end uses of nuclear energy are diverse and include electricity, process heat, water desalination, district heating, and possibly future hydrogen production for transportation and energy storage uses. Components of the nuclear fuel cycle include front end components such as uranium mining, conversion and enrichment, fuel fabrication, and the reactor component. Back end of the fuel cycle include used fuel coming out the reactor, used fuel temporary and permanent storage, and fuel reprocessing. Combined with those components there

39

Nuclear & Uranium - U.S. Energy Information Administration (EIA) - U.S.  

Gasoline and Diesel Fuel Update (EIA)

Nuclear & Uranium Nuclear & Uranium Glossary › FAQS › Overview Data Summary Uranium & Nuclear Fuel Nuclear Power Plants Radioactive Waste International All Nuclear Data Reports Analysis & Projections Most Requested Nuclear Plants and Reactors Projections Uranium All Reports Uranium Mill Sites Under the UMTRA Project Remediation of UMTRCA Title I Uranium Mill Sites Under the UMTRA Project Summary Table: Uranium Ore Processed, Disposal Cell Material, and Cost for Remediation as of December 31, 1999 Uranium Ore Processed Remediation Project Cost Remediation Project (Mill Site Name, State) Ore (Million Short Tons) Uranium Production (Million Pounds U3O8) Disposal Cell Remediated Material Volume (Million Cubic Yards) Total Cost A (Thousand U.S. Dollars)02/09 Per Pound Produced (Dollars per Pound U3O8) Per Unit of Remediated Material

40

Pyroprocessing of fast flux test facility nuclear fuel  

SciTech Connect (OSTI)

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

Westphal, B.R.; Wurth, L.A.; Fredrickson, G.L.; Galbreth, G.G.; Vaden, D.; Elliott, M.D.; Price, J.C.; Honeyfield, E.M.; Patterson, M.N. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID, 83415 (United States)

2013-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Extraction of uranium from spent fuels using liquefied gases  

SciTech Connect (OSTI)

For reprocessing of spent nuclear fuels, a novel method to extract actinides from spent fuel using highly compressed gases, nitrogen dioxide and carbon dioxide was proposed. As a fundamental study, the nitrate conversion with liquefied nitrogen dioxide and the nitrate extraction with supercritical carbon dioxide were demonstrated by using uranium dioxide powder, uranyl nitrate and tri-n-butylphosphate complex in the present study. (authors)

Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi [EcoTopia Science Institute, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, 464-8603 (Japan)

2007-07-01T23:59:59.000Z

42

EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing...  

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

Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing U.S.-Origin...

43

RADIATION DOSE ASPECTS IN THE HANDLING OF EMERGING NUCLEAR FUELS  

Science Journals Connector (OSTI)

......Prot. (2008) 28:161. 15 NUREG. Standard review plan for the review of an application for a Mixed Oxide (MOX) fuel...fabrication facility. (2000) NUREG-1718, US Nuclear Regulatory Commission. 16 IAEA. Safety of uranium fuel fabrication......

G. Nicolaou

2014-02-01T23:59:59.000Z

44

Nuclear & Uranium - Data - U.S. Energy Information Administration (EIA)  

Gasoline and Diesel Fuel Update (EIA)

Find statistics on nuclear operable units, nuclear electricity net Find statistics on nuclear operable units, nuclear electricity net generation, nuclear share of electricity net generation, and capacity factor. + EXPAND ALL Summary Additional Formats Nuclear Overview: PDF CSV XLS Monthly statistics on nuclear operable units, nuclear electricity net generation, nuclear share of electricity net generation, and capacity factor. PDFXLS Annual statistics on nuclear generating units, power plants operations, and uranium. › Nuclear Generating Units, 1955-2010 › PDF XLS Nuclear Power Plant Operations, 1957-2010 › PDF XLS Uranium Overview, 1949-2010 › PDF XLS Uranium & Nuclear Fuel Additional Formats U.S. Uranium Reserves Estimates › Release Date: July 2010 The U.S. Energy Information Administration (EIA) has updated its estimates of uranium reserves for year-end 2008. This represents the first revision of the estimates since 2004. PDF

45

Fuel Cycle Science & Technology | Nuclear Science | ORNL  

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

Advanced Fuel Cycle Systems Radiochemical Separation & Processing Recycle & Waste Management Uranium Enrichment Used Nuclear Fuel Storage, Transportation, and Disposal Fusion Nuclear Science Isotope Development and Production Nuclear Security Science & Technology Nuclear Systems Modeling, Simulation & Validation Nuclear Systems Technology Reactor Technology Nuclear Science Home | Science & Discovery | Nuclear Science | Research Areas | Fuel Cycle Science & Technology SHARE Fuel Cycle Science and Technology The ORNL expertise and experience across the entire nuclear fuel cycle is underpinned by extensive facilities and a comprehensive modeling and simulation capability ORNL supports the understanding, development, evaluation and deployment of

46

6 Nuclear Fuel Designs  

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

Message from the Director Message from the Director 2 Nuclear Power & Researrh Reactors 3 Discovery of Promethium 4 Nuclear Isotopes 4 Nuclear Medicine 5 Nuclear Fuel Processes & Software 6 Nuclear Fuel Designs 6 Nuclear Safety 7 Nuclear Desalination 7 Nuclear Nonproliferation 8 Neutron Scattering 9 Semiconductors & Superconductors 10 lon-Implanted Joints 10 Environmental Impact Analyses 11 Environmental Quality 12 Space Exploration 12 Graphite & Carbon Products 13 Advanced Materials: Alloys 14 Advanced Materials: Ceramics 15 Biological Systems 16 Biological Systems 17 Computational Biology 18 Biomedical Technologies 19 Intelligent Machines 20 Health Physics & Radiation Dosimetry 21 Radiation Shielding 21 Information Centers 22 Energy Efficiency: Cooling & Heating

47

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

DOE Patents [OSTI]

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuels is disclosed using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuels, two cathodes and electrical power means connected to the anode basket, cathodes and lower molten cadmium pool for providing electrical power to the cell. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then purified uranium is electrolytically transported and deposited on a first molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on a second cathode. 3 figs.

Ackerman, J.P.; Miller, W.E.

1987-11-05T23:59:59.000Z

48

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

DOE Patents [OSTI]

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

Ackerman, John P. (Downers Grove, IL); Miller, William E. (Naperville, IL)

1989-01-01T23:59:59.000Z

49

World nuclear fuel cycle requirements 1990  

SciTech Connect (OSTI)

This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under three nuclear supply scenarios. Two of these scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries with free market economies (FME countries). A No New Orders scenario is presented only for the United States. These nuclear supply scenarios are described in Commercial Nuclear Power 1990: Prospects for the United States and the World (DOE/EIA-0438(90)). This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the FME projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2030 for the Lower and Upper Reference cases and through 2040, the last year in which spent fuel is discharged, for the No New Orders case. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management.

Not Available

1990-10-26T23:59:59.000Z

50

Oxidation Protection of Uranium Nitride Fuel using Liquid Phase Sintering  

SciTech Connect (OSTI)

Two methods are proposed to increase the oxidation resistance of uranium nitride (UN) nuclear fuel. These paths are: (1) Addition of USi{sub x} (e.g. U3Si2) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with various compounds (followed by densification via Spark Plasma Sintering or Liquid Phase Sintering) that will greatly increase oxidation resistance. The advantages (high thermal conductivity, very high melting point, and high density) of nitride fuel have long been recognized. The sodium cooled BR-10 reactor in Russia operated for 18 years on uranium nitride fuel (UN was used as the driver fuel for two core loads). However, the potential advantages (large power up-grade, increased cycle lengths, possible high burn-ups) as a Light Water Reactor (LWR) fuel are offset by uranium nitride's extremely low oxidation resistance (UN powders oxidize in air and UN pellets decompose in hot water). Innovative research is proposed to solve this problem and thereby provide an accident tolerant LWR fuel that would resist water leaks and high temperature steam oxidation/spalling during an accident. It is proposed that we investigate two methods to increase the oxidation resistance of UN: (1) Addition of USi{sub x} (e.g. U{sub 3}Si{sub 2}) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with compounds (followed by densification via Spark Plasma Sintering) that will greatly increase oxidation resistance.

Dr. Paul A. Lessing

2012-03-01T23:59:59.000Z

51

Uranium mononitride as a potential commercial LWR fuel  

SciTech Connect (OSTI)

This paper evaluated uranium mononitride (UN) as a potential replacement for 5% enriched UO{sub 2} fuel in Generation III and III+ commercial light water reactors (LWRs). Significant improvement in LWR performance depends on developing and implementing changes in the nuclear fuel used in these reactors. Compared to UO{sub 2}, UN offers several advantages such as higher uranium loading and better thermal conductivity. In this paper, the thermal safety margin of UN was evaluated at both normal and accident conditions using a readily available coupled CFD model developed for the US DOE CASL program. One of the prime technical challenges in utilization of UN as LWR fuel is the water compatibility because pure phase UN is not stable in water at 350 deg. C. The water corrosion resistance of UN and the corrosion mechanism were reviewed and mitigation methods were proposed. (authors)

Xu, P.; Yan, J.; Lahoda, E. J.; Ray, S. [Westinghouse Electric Company, LLC, 5801 Bluff Rd, Columbia, SC 29209 (United States)

2012-07-01T23:59:59.000Z

52

Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel  

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

Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel Production: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel ... Fact Sheet Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel

53

Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel  

National Nuclear Security Administration (NNSA)

Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel Production: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel ... Fact Sheet Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel

54

Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel  

SciTech Connect (OSTI)

Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR`s uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ``hot segment`` analysis of narrow axial regions along the plate and ``hot streak`` analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about {minus}7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square ({chi}{sup 2}) test for goodness of fit to normal distributions was not satisfied.

Blumenfeld, P.E.

1995-08-01T23:59:59.000Z

55

Uranium chloride extraction of transuranium elements from LWR fuel  

DOE Patents [OSTI]

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

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

1992-08-25T23:59:59.000Z

56

World nuclear capacity and fuel cycle requirements, November 1993  

SciTech Connect (OSTI)

This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy`s activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration`s annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment.

Not Available

1993-11-30T23:59:59.000Z

57

Methods for making a porous nuclear fuel element  

SciTech Connect (OSTI)

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Youchison, Dennis L; Williams, Brian E; Benander, Robert E

2014-12-30T23:59:59.000Z

58

Colloids generation from metallic uranium fuel  

SciTech Connect (OSTI)

The possibility of colloid generation from spent fuel in an unsaturated environment has significant implications for storage of these fuels in the proposed repository at Yucca Mountain. Because colloids can act as a transport medium for sparingly soluble radionuclides, it might be possible for colloid-associated radionuclides to migrate large distances underground and present a human health concern. This study examines the nature of colloidal materials produced during corrosion of metallic uranium fuel in simulated groundwater at elevated temperature in an unsaturated environment. Colloidal analyses of the leachates from these corrosion tests were performed using dynamic light scattering and transmission electron microscopy. Results from both techniques indicate a bimodal distribution of small discrete particles and aggregates of the small particles. The average diameters of the small, discrete colloids are {approximately}3--12 nm, and the large aggregates have average diameters of {approximately}100--200 nm. X-ray diffraction of the solids from these tests indicates a mineral composition of uranium oxide or uranium oxy-hydroxide.

Metz, C.; Fortner, J.; Goldberg, M.; Shelton-Davis, C.

2000-07-20T23:59:59.000Z

59

Description of the Canadian Particulate-Fill WastePackage (WP) System for Spent-Nuclear Fuel (SNF) and its Applicability to Ligh-Water Reactor SNF WPS with Depleted Uranium-Dioxide Fill  

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

3502 3502 Chemical Technology Division DESCRIPTION OF THE CANADIAN PARTICULATE-FILL WASTE-PACKAGE (WP) SYSTEM FOR SPENT-NUCLEAR FUEL(SNF) AND ITS APPLICABILITY TO LIGHT- WATER REACTOR SNF WPS WITH DEPLETED URANIUM-DIOXIDE FILL Charles W. Forsberg Oak Ridge National Laboratory * P.O. Box 2008 Oak Ridge, Tennessee 37831-6180 Tel: (423) 574-6783 Fax: (423) 574-9512 Email: forsbergcw@ornl.gov October 20, 1997 _________________________ Managed by Lockheed Martin Energy Research Corp. under contract DE-AC05-96OR22464 for the * U.S. Department of Energy. iii CONTENTS LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v LIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

60

Technology Insights and Perspectives for Nuclear Fuel Cycle Concepts  

SciTech Connect (OSTI)

The following report provides a rich resource of information for exploring fuel cycle characteristics. The most noteworthy trends can be traced back to the utilization efficiency of natural uranium resources. By definition, complete uranium utilization occurs only when all of the natural uranium resource can be introduced into the nuclear reactor long enough for all of it to undergo fission. Achieving near complete uranium utilization requires technologies that can achieve full recycle or at least nearly full recycle of the initial natural uranium consumed from the Earth. Greater than 99% of all natural uranium is fertile, and thus is not conducive to fission. This fact requires the fuel cycle to convert large quantities of non-fissile material into fissile transuranics. Step increases in waste benefits are closely related to the step increase in uranium utilization going from non-breeding fuel cycles to breeding fuel cycles. The amount of mass requiring a disposal path is tightly coupled to the quantity of actinides in the waste stream. Complete uranium utilization by definition means that zero (practically, near zero) actinide mass is present in the waste stream. Therefore, fuel cycles with complete (uranium and transuranic) recycle discharge predominately fission products with some actinide process losses. Fuel cycles without complete recycle discharge a much more massive waste stream because only a fraction of the initial actinide mass is burned prior to disposal. In a nuclear growth scenario, the relevant acceptable frequency for core damage events in nuclear reactors is inversely proportional to the number of reactors deployed in a fuel cycle. For ten times the reactors in a fleet, it should be expected that the fleet-average core damage frequency be decreased by a factor of ten. The relevant proliferation resistance of a fuel cycle system is enhanced with: decreasing reliance on domestic fuel cycle services, decreasing adaptability for technology misuse, enablement of material accountability, and decreasing material attractiveness.

S. Bays; S. Piet; N. Soelberg; M. Lineberry; B. Dixon

2010-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Computational Design of Advanced Nuclear Fuels  

SciTech Connect (OSTI)

The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

2014-06-03T23:59:59.000Z

62

Uranium Mining and Enrichment  

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

Overview Presentation » Uranium Mining and Enrichment Overview Presentation » Uranium Mining and Enrichment Uranium Mining and Enrichment Uranium is a radioactive element that occurs naturally in the earth's surface. Uranium is used as a fuel for nuclear reactors. Uranium-bearing ores are mined, and the uranium is processed to make reactor fuel. In nature, uranium atoms exist in several forms called isotopes - primarily uranium-238, or U-238, and uranium-235, or U-235. In a typical sample of natural uranium, most of the mass (99.3%) would consist of atoms of U-238, and a very small portion of the total mass (0.7%) would consist of atoms of U-235. Uranium Isotopes Isotopes of Uranium Using uranium as a fuel in the types of nuclear reactors common in the United States requires that the uranium be enriched so that the percentage of U-235 is increased, typically to 3 to 5%.

63

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

SciTech Connect (OSTI)

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

Heidet, F.; Kim, T.; Grandy, C. (Nuclear Engineering Division)

2012-07-30T23:59:59.000Z

64

Nuclear fuel cycle information workshop  

SciTech Connect (OSTI)

This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work; second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity; and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US.

Not Available

1983-01-01T23:59:59.000Z

65

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors  

DOE Patents [OSTI]

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03T23:59:59.000Z

66

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors  

DOE Patents [OSTI]

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Youchison, Dennis L. (Albuquerque, NM); Williams, Brian E. (Pacoima, CA); Benander, Robert E. (Pacoima, CA)

2011-03-01T23:59:59.000Z

67

Nuclear Spent Fuel Program Drivers  

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

was created to plan and coordinate the management of Department of Energy-owned spent nuclear fuel. It was established as a result of a 1992 decision to stop spent nuclear fuel...

68

Power Surge: Uranium alloy fuel for TerraPower | Y-12 National Security  

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

Power Surge: Uranium alloy ... Power Surge: Uranium alloy ... Power Surge: Uranium alloy fuel for TerraPower Posted: July 18, 2012 - 9:45am | Y-12 Report | Volume 9, Issue 1 | 2012 Since 2010, Y-12 has provided TerraPower with technical support in the fabrication methods for uranium alloy fuel to be used in a new traveling wave nuclear reactor that can run for more than 30 years without refueling. Image of reactor power concept, used with permission of TerraPower, LLC. Y-12's nuclear expertise, expanding since the site's integral role in the Manhattan Project, is positioning the Y-12 Complex at the forefront of what Sen. Lamar Alexander repeatedly asserts is needed - "a new Manhattan Project for clean energy independence." TerraPower, a private company backed by Microsoft founder Bill Gates, is

69

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

70

LMFBR operation in the nuclear cycle without fuel reprocessing  

SciTech Connect (OSTI)

Substantiation is given to expediency of investigation of nuclear power (NP) development with fast reactors cooled by lead-bismuth alloy operating during extended time in the open nuclear fuel cycle with slightly enriched or depleted uranium make-up. 9 refs., 1 fig., 6 tabs.

Toshinsky, S.I. [Institute of Physics and Power Engineering, Kaluga (Russian Federation)

1997-12-01T23:59:59.000Z

71

Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site  

SciTech Connect (OSTI)

The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

Pinkston, Daniel [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL

2010-10-01T23:59:59.000Z

72

International Atomic Energy Agency support of research reactor highly enriched uranium to low enriched uranium fuel conversion projects  

SciTech Connect (OSTI)

The IAEA has been involved for more than twenty years in supporting international nuclear non- proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly assisted efforts to convert research reactors from HEU to LEU fuel. HEU to LEU fuel conversion projects differ significantly depending on several factors including the design of the reactor and fuel, technical needs of the member state, local nuclear infrastructure, and available resources. To support such diverse endeavours, the IAEA tailors each project to address the relevant constraints. This paper presents the different approaches taken by the IAEA to address the diverse challenges involved in research reactor HEU to LEU fuel conversion projects. Examples of conversion related projects in different Member States are fully detailed. (author)

Bradley, E.; Adelfang, P.; Goldman, I.N. [Research Reactors Unit, Division of Nuclear Fuel Cycle and Waste Technology, International Atomic Energy Agency, Wagramer Strasse 5, P.O. Box 100, A-1400 Vienna (Austria)

2008-07-15T23:59:59.000Z

73

Swelling-resistant nuclear fuel  

DOE Patents [OSTI]

A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

Arsenlis, Athanasios (Hayward, CA); Satcher, Jr., Joe (Patterson, CA); Kucheyev, Sergei O. (Oakland, CA)

2011-12-27T23:59:59.000Z

74

Spent Nuclear Fuel Fact Sheets  

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

management needs. By coordinating common needs for research, technology development, and testing programs, the National Spent Nuclear Fuel Program is achieving cost efficiencies...

75

Uranium-plutonium-neptunium fuel cycle to produce isotopically denatured plutonium  

SciTech Connect (OSTI)

In view of the considerable amount of /sup 237/ Np produced as a by-product in nuclear power reactors, possible utilization of this nuclide in the nuclear fuel cycle has been studied. In particular, the performance of a gas-cooled fast breeder reactor as a neptunium burner was assessed. A strategy was developed and mass flows were computed for a denatured plutonium LWR strategy using uranium, plutonium and neptunium recycling. 10 refs.

Wydler, P.; Heer, W.; Stiller, P.; Wenger, H.U.

1980-06-01T23:59:59.000Z

76

Irradiation behavior of miniature experimental uranium silicide fuel plates  

SciTech Connect (OSTI)

Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 10/sup 20/ cm/sup -3/, far short of the approximately 20 x 10/sup 20/ cm/sup -3/ goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix.

Hofman, G.L.; Neimark, L.A.; Mattas, R.F.

1983-01-01T23:59:59.000Z

77

FABRICATION OF URANIUM OXYCARBIDE KERNELS AND COMPACTS FOR HTR FUEL  

SciTech Connect (OSTI)

As part of the program to demonstrate tristructural isotropic (TRISO)-coated fuel for the Next Generation Nuclear Plant (NGNP), Advanced Gas Reactor (AGR) fuel is being irradiation tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). This testing has led to improved kernel fabrication techniques, the formation of TRISO fuel particles, and upgrades to the overcoating, compaction, and heat treatment processes. Combined, these improvements provide a fuel manufacturing process that meets the stringent requirements associated with testing in the AGR experimentation program. Researchers at Idaho National Laboratory (INL) are working in conjunction with a team from Babcock and Wilcox (B&W) and Oak Ridge National Laboratory (ORNL) to (a) improve the quality of uranium oxycarbide (UCO) fuel kernels, (b) deposit TRISO layers to produce a fuel that meets or exceeds the standard developed by German researches in the 1980s, and (c) develop a process to overcoat TRISO particles with the same matrix material, but applies it with water using equipment previously and successfully employed in the pharmaceutical industry. A primary goal of this work is to simplify the process, making it more robust and repeatable while relying less on operator technique than prior overcoating efforts. A secondary goal is to improve first-pass yields to greater than 95% through the use of established technology and equipment. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 to November 2009. The AGR-1 fuel was designed to closely replicate many of the properties of German TRISO-coated particles, thought to be important for good fuel performance. No release of gaseous fission product, indicative of particle coating failure, was detected in the nearly 3-year irradiation to a peak burn up of 19.6% at a time-average temperature of 1038–1121°C. Before fabricating AGR-2 fuel, each fabrication process was improved and changed. Changes to the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a 6-inch diameter coater using a charge size about 21-times that of the 2-inch diameter coater used to coat AGR-1 particles. The compacting process was changed to increase matrix density and throughput by increasing the temperature and pressure of pressing and using a different type of press. AGR-2 fuel began irradiation in the ATR in late spring 2010.

Dr. Jeffrey A. Phillips; Eric L. Shaber; Scott G. Nagley

2012-10-01T23:59:59.000Z

78

Possibility of nuclear pumped laser experiment using low enriched uranium  

SciTech Connect (OSTI)

Possibility to perform experiments for nuclear pumped laser oscillation by using low enriched uranium is investigated. Kinetic analyses are performed for two types of reactor design, one is using highly enriched uranium and the other is using low enriched uranium. The reactor design is based on the experiment reactor in IPPE. The results show the oscillation of nuclear pumped laser in the case of low enriched uranium reactor is also possible. The use of low enriched uranium in the experiment will make experiment easier.

Obara, Toru; Takezawa, Hiroki [Center for Research into Innovative Nuclear Energy Systems Tokyo Institute of Technology 2-12-1-N1-19, Ookayama Meguro-ku, Tokyo 152-8550 (Japan)

2012-06-06T23:59:59.000Z

79

Energy Return on Investment from Recycling Nuclear Fuel  

SciTech Connect (OSTI)

This report presents an evaluation of the Energy Return on Investment (EROI) from recycling an initial batch of 800 t/y of used nuclear fuel (UNF) through a Recycle Center under a number of different fuel cycle scenarios. The study assumed that apart from the original 800 t of UNF only depleted uranium was available as a feed. Therefore for each subsequent scenario only fuel that was derived from the previous fuel cycle scenario was considered. The scenarios represent a good cross section of the options available and the results contained in this paper and associated appendices will allow for other fuel cycle options to be considered.

None

2011-08-17T23:59:59.000Z

80

Characteristics of a Mixed Thorium-Uranium Dioxide High-Burnup Fuel  

SciTech Connect (OSTI)

Future nuclear fuels must satisfy three sets of requirements: longer times between refueling; concerns for weapons proliferation; and development of a spent fuel form more suitable for direct geologic disposal. This project has investigated a fuel consisting of mixed thorium and uranium dioxide to satisfy these requirements. Results using the SCALE 4.3 code system indicated that the mixed Th-U fuel could be burned to 72 MWD/kg or 100 MWD/kg using 25% of 35% UO2 respectively. The uranium remained below 20% total fissile fraction throughout the cycle, making it unusable for weapons. Total plutonium production per MWD was a factor of 4.5 less in the Th-U fuel than in the conventional fuel; Pu-239 production per MWD was a factor of 6.5 less; and the plutonium produced was high in Pu-238, leading to a decay heat 5 times greater than that from plutonium derived from conventional fuel and 40 times greater than weapons grade plutonium. High decay heat would require active cooling of any crude weapon, lest the components surrounding the plutonium be melted. Spontaneous neutron production for plutonium from Th-U fuel was 2.3 times greater than that from conventional fuel and 15 times greater than that from weapons grade plutonium. High spontaneous neutron production drastically limits the probable yield of a crude weapon. Because ThO2 is the highest oxide of thorium, while UO2 can be oxidized further to U3O8, ThO2-UO2 fuel may be a superior wasteform if the spent fuel is ever to be exposed to oxygenated water. Even if the cost of fabricating the mixed Th-U fuel is $100/kg greater, the cost of the Th-U fuel is 13% to 15% less than that of the fuels using uranium only.

J. S. Herring; P. E. MacDonald

1999-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Characteristics of a Mixed Thorium - Uranium Dioxide High-Burnup Fuel  

SciTech Connect (OSTI)

Future nuclear fuel must satisfy three sets of requirements: longer times between refueling; concerns for weapons proliferation; and development of a spent fuel form more suitable for direct geologic disposal. This project has investigated a fuel consisting of mixed thorium and uranium dioxide to satisfy these requirements. Results using the SCALE 4.3 code system indicated that the mixed Th-U fuel could be burned to 72 MWD/kg or 100 MWD/kg using 25% and 35% UO2 respectively. The uranium remained below 20 % total fissile fraction throughout the cycle, making it unusable for weapons. Total plutonium production per MWD was a factor of 4.5 less in the Th-U fuel than in the conventional fuel; Pu-239 production per MWD was a factor of 6.5 less; and the plutonium produced was high in Pu-238, leading to a decay heat 5 times greater than that from plutonium derived from conventional fuel and 40 times greater than weapons grade plutonium. High decay heat would require active cooling of any crude weapon, lest the components surrounding the plutonium be melted. Spontaneous neutron production for plutonium from Th-U fuel was 2.3 times greater than that from conventional fuel and 15 times greater than that from weapons grade plutonium. High spontaneous neutron production drastically limits the probable yield of a crude weapon. Because ThO2 is the highest oxide of thorium, while UO2 can be oxidized further to U3O8, ThO2- UO2 fuel may be a superior wasteform if the spent fuel is ever to be exposed to oxygenated water. Even if the cost of fabricating the mixed Th-U fuel is $100/kg greater, the cost of the Th-U fuel is 13% to 25% less than that of the fuels using uranium only.

Herring, James Stephen; Mac Donald, Philip Elsworth

1999-06-01T23:59:59.000Z

82

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

E-Print Network [OSTI]

Level Scheme Library Depleted Uranium Evaluated Nuclear Datafuel and the 238 U in depleted uranium (DU) was used as a

Ludewigt, Bernhard A

2011-01-01T23:59:59.000Z

83

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

SciTech Connect (OSTI)

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

Sean M. McDeavitt

2011-04-29T23:59:59.000Z

84

Pyrochemical processing of DOE spent nuclear fuel  

SciTech Connect (OSTI)

A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or {open_quotes}pyroprocessing,{close_quotes} provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (>99.9%) separation of transuranics. The resultant waste forms from the pyroprocess, are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory.

Laidler, J.J.

1995-02-01T23:59:59.000Z

85

Advances in Metallic Nuclear Fuel  

Science Journals Connector (OSTI)

Metallic nuclear fuels have generated renewed interest for advanced ... operations is excellent. Ongoing irradiation tests in Argonne-West’s Idaho-based Experimental Breeder Reactor ... fast reactor (IFR) concept...

B. R. Seidel; L. C. Walters; Y. I. Chang

1987-04-01T23:59:59.000Z

86

Evaluating Environmental, Health and Safety Impacts from Two Nuclear Fuel Cycles: A Comparative Analysis of Once-Through Uranium Use and Plutonium Recycle in Light Water Reactors.  

E-Print Network [OSTI]

??Prioritizing the finite resources available to advance research, development and demonstration of the nuclear industry requires a comprehensive evaluation of potential advanced nuclear technologies to… (more)

Smith, Bethany Lee

2014-01-01T23:59:59.000Z

87

The gas centrifuge and nuclear weapons proliferation  

SciTech Connect (OSTI)

Uranium enrichment by centrifugation is the basis for the quick and efficient production of nuclear fuel-or nuclear weapons.

Wood, Houston G. [Mechanical and Aerospace Engineering Department, University of Virginia, Charlottesville, Virginia (United States); Glaser, Alexander [Woodrow Wilson School of Public and International Affairs, Program on Science, Technology and Environmental Policy, Princeton University, Princeton, New Jersey (United States); Kemp, R. Scott [Nuclear Science and Engineering Department, Massachusetts Institute of Technology, Cambridge, Massachusetts (United States)

2014-05-09T23:59:59.000Z

88

Nuclear Fuel Cycle & Vulnerabilities  

SciTech Connect (OSTI)

The objective of safeguards is the timely detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. The safeguards system should be designed to provide credible assurances that there has been no diversion of declared nuclear material and no undeclared nuclear material and activities.

Boyer, Brian D. [Los Alamos National Laboratory

2012-06-18T23:59:59.000Z

89

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

90

Advanced Nuclear Fuel | Y-12 National Security Complex  

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

operate; lowers proliferation risks by reducing the need for enriched uranium; converts depleted uranium to usable fuel as it operates; uses liquid sodium as a coolant, which is...

91

Sustainability Features of Nuclear Fuel Cycle Options  

E-Print Network [OSTI]

The nuclear fuel cycle is the series of stages that nuclear fuel materials go through in a cradle to grave framework. The Once Through Cycle (OTC) is the current fuel cycle implemented in the United States; in which an ...

Passerini, Stefano

92

Empirical modeling of uranium nitride fuels  

E-Print Network [OSTI]

SD Fuel swelling ( volume % ) Fission gas release (% ) Area average fuel temperature at the peak axial location Fuel burnup Fuel density Smear density The empirical fits shown above were produced using a least squares fit program with data... rejected due to a demonstrated lack of stability. The fuel swelling and fission gas release values predicted by the nonlinear correlations show fair agreement with the two experimental pins from the SP-1 irradiation test . Additionally, the trends...

Brozak, Daniel Edward

2012-06-07T23:59:59.000Z

93

Nuclear fuel recycling in 4 minutes | Argonne National Laboratory  

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

Nuclear fuel recycling in 4 minutes Share Topic Energy Energy sources Nuclear energy Nuclear fuel cycle Reactors...

94

Production and Characterization of Monodisperse Plutonium, Uranium, and Mixed Uranium?Plutonium Particles for Nuclear Safeguard Applications  

Science Journals Connector (OSTI)

Production and Characterization of Monodisperse Plutonium, Uranium, and Mixed Uranium?Plutonium Particles for Nuclear Safeguard Applications ... In order to prevent nuclear proliferation, the isotopic analysis of uranium and plutonium microparticles has strengthened the means in international safeguards for detecting undeclared nuclear activities. ...

Y. Ranebo; N. Niagolova; N. Erdmann; M. Eriksson; G. Tamborini; M. Betti

2010-04-23T23:59:59.000Z

95

Nuclear reactors and the nuclear fuel cycle  

SciTech Connect (OSTI)

According to the author, the first sustained nuclear fission chain reaction was not at the University of Chicago, but at the Oklo site in the African country of Gabon. Proof of this phenomenon is provided by mass spectrometric and analytical chemical measurements by French scientists. The U.S. experience in developing power-producing reactors and their related fuel and fuel cycles is discussed.

Pearlman, H

1989-11-01T23:59:59.000Z

96

Observation of the uranium 235 nuclear magnetic resonance signal (*)  

E-Print Network [OSTI]

before. We report here the first NMR observation of 23SU. The uranium hexafluoride has been chosenL-1017 Observation of the uranium 235 nuclear magnetic resonance signal (*) H. Le Bail, C. Chachaty signal de résonance magnétique nucléaire de l'isotope 235 de l'uranium est présentée. Elle a été

Paris-Sud XI, Université de

97

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

98

Oak Ridge, Tenn. Selected as Uranium Enrichment Site | National Nuclear  

National Nuclear Security Administration (NNSA)

Oak Ridge, Tenn. Selected as Uranium Enrichment Site | National Nuclear Oak Ridge, Tenn. Selected as Uranium Enrichment Site | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Oak Ridge, Tenn. Selected as Uranium Enrichment Site Oak Ridge, Tenn. Selected as Uranium Enrichment Site September 19, 1942 Oak Ridge, TN

99

Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium  

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

Nuclear Materials & Waste » Nuclear Materials & Waste » Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium and Uranium-233 Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium and Uranium-233 105-K building houses the K-Area Material Storage (KAMS) facility, designated for the consolidated storage of surplus plutonium at Savannah River Site pending disposition. The plutonium shipped to KAMS is sealed inside a welded 3013 containers that are nested in 9975 shipping containers. 105-K building houses the K-Area Material Storage (KAMS) facility, designated for the consolidated storage of surplus plutonium at Savannah River Site pending disposition. The plutonium shipped to KAMS is sealed inside a welded 3013 containers that are nested in 9975 shipping

100

Activities Related to Storage of Spent Nuclear Fuel | Department...  

Office of Environmental Management (EM)

Activities Related to Storage of Spent Nuclear Fuel Activities Related to Storage of Spent Nuclear Fuel Activities Related to Storage of Spent Nuclear Fuel More Documents &...

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

of hydride fueled BWRs. Nuclear Engineering and Design, 239:Fueled PWR Cores. Nuclear Engineering and Design, 239:1489–Hydride Fueled LWRs. Nuclear Engineering and Design, 239:

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

102

Nuclear & Uranium - Analysis & Projections - U.S. Energy Information  

Gasoline and Diesel Fuel Update (EIA)

Most Requested Most Requested Change category... Most Requested Nuclear Plants and Reactors Projections Uranium All Reports Filter by: All Data Analysis Projections Weekly Reports Today in Energy - Nuclear Short, timely articles with graphs about recent nuclear energy issues and trends Monthly Reports Monthly Energy Review - Nuclear Section Released: November 25, 2013 Recent statistics on nuclear electricity capacity, generation, and number of operable nuclear reactors. Electricity Monthly Update Released: November 22, 2013 Provides analysis and of the highlights of the data included in the Electric Power Monthly publication and presents tables of electricity generation, fuel consumption for generation, fossil fuel stocks, and average retail sales and prices of electricity. The EMU is published at the

103

Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report  

SciTech Connect (OSTI)

The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

William Anderson; James Tulenko; Bradley Rearden; Gary Harms

2008-09-11T23:59:59.000Z

104

RADIATION DOSE ASPECTS IN THE HANDLING OF EMERGING NUCLEAR FUELS  

Science Journals Connector (OSTI)

......transmutation in LMFBR, and uranium (U) matrix fuels...161. 15 NUREG. Standard review plan for the review of an application...16 IAEA. Safety of uranium fuel fabrication facilities...2010) IAEA Safety Standards Series No. SSG-6......

G. Nicolaou

2014-02-01T23:59:59.000Z

105

Corrosion of Spent Nuclear Fuel: The Long-Term Assessment  

SciTech Connect (OSTI)

Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

Rodney C. Ewing

2004-10-07T23:59:59.000Z

106

Engineering of Deinococcus radiodurans R1 for Bioprecipitation of Uranium from Dilute Nuclear Waste  

Science Journals Connector (OSTI)

...radiodurans R1 for Bioprecipitation of Uranium from Dilute Nuclear Waste Published ahead of print on 20 October 2006. Deepti...situ approach to biorecovery of uranium from dilute nuclear waste. Nuclear waste contains a variety of heavy metals, radionuclides...

Deepti Appukuttan; Amara Sambasiva Rao; Shree Kumar Apte

2006-10-20T23:59:59.000Z

107

Microscopic Examination of a Corrosion Front in Spent Nuclear Fuel  

SciTech Connect (OSTI)

Spent uranium oxide nuclear fuel hosts a variety of trace chemical constituents, many of which must be sequestered from the biosphere during fuel storage and disposal. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within spent uranium oxide nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the plutonium and neptunium in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO{sub 2} matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl U(VI)O{sub 2}{sup 2+} mineral assemblage that is depleted in plutonium and neptunium relative to the parent fuel. At the corrosion front interface between intact fuel and the uranyl-mineral corrosion layer, we find evidence of a thin ({approx}20 micrometer) layer that is enriched in plutonium and neptunium within a predominantly U{sup 4+} environment. Available data for the standard reduction potentials for NpO{sup 2+}/Np{sup 4+} and UO{sub 2}{sup 2+}/U{sup 4+} couples indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potentials of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. Neptunium is an important radionuclide in dose contribution according to performance assessment models of the proposed U. S. repository at Yucca Mountain, Nevada. A scientific understanding of how the UO{sub 2} matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of neptunium is needed to predict its behavior at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel [1]. In the immediate vicinity of the spent fuel's surface the redox and nucleation behavior is likely to promote/enhance nucleation of NpO{sub 2} and Np{sub 2}O{sub 5}. Alternatively, Np may be incorporated into uranyl (UO{sub 2}{sup 2+}) alteration phases [2]. In some cases, less-soluble elements such as plutonium will be enriched near the surface of the corroding fuel [3]. We have used focused synchrotron x-rays from the MRCAT beam line at the Advanced Photon Source (APS) at Argonne National Lab to examine a specimen of spent nuclear fuel that had been subject to 10 years of corrosion testing in an environment of humid air and dripping groundwater at 90 C [4]. We find evidence of a region, approximately 20 microns in thickness, enriched in plutonium and neptunium at the corrosion front that exists between the uranyl silicate alteration mineral rind and the unaltered uranium oxide fuel (Figures 1 and 2). The uranyl silicate is itself found to be depleted in these transuranic elements relative to their abundance relative to uranium in the parent fuel. This suggests a low mobility of these components owing to a resistance to oxidize further in the presence of a UO{sub 2}{sup 2+}/U{sup 4+} couple [5].

J.A> Fortner; A.J. Kropf; R.J. Finch; J.C. Cunnane

2006-06-20T23:59:59.000Z

108

Characterization plan for Hanford spent nuclear fuel  

SciTech Connect (OSTI)

Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF.

Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

1994-12-01T23:59:59.000Z

109

Spent nuclear fuel reprocessing modeling  

SciTech Connect (OSTI)

The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

Tretyakova, S.; Shmidt, O.; Podymova, T.; Shadrin, A.; Tkachenko, V. [Bochvar Institute, 5 Rogova str., Moscow 123098 (Russian Federation); Makeyeva, I.; Tkachenko, V.; Verbitskaya, O.; Schultz, O.; Peshkichev, I. [Russian Federal Nuclear Center - VNIITF E.I. Zababakhin, p.o.box 245, Snezhinsk, 456770 (Russian Federation)

2013-07-01T23:59:59.000Z

110

Sustainability Considerations in Spent Light-water Nuclear Fuel Retrievability  

SciTech Connect (OSTI)

This paper examines long-term cost differences between two competing Light Water Reactor (LWR) fuels: Uranium Oxide (UOX) and Mixed Uranium Oxide-Plutonium Oxide (MOX). Since these costs are calculated on a life-cycle basis, expected savings from lower future MOX fuel prices can be used to value the option of substituting MOX for UOX, including the value of maintaining access to the used UOX fuel that could be reprocessed to make MOX. The two most influential cost drivers are the price of natural uranium and the cost of reprocessing. Significant and sustained reductions in reprocessing costs and/or sustained increases in uranium prices are required to give positive value to the retrievability of Spent Nuclear Fuel. While this option has positive economic value, it might not be exercised for 50 to 200 years. Therefore, there are many years for a program during which reprocessing technology can be researched, developed, demonstrated, and deployed. Further research is required to determine whether the cost of such a program would yield positive net present value and/or increases the sustainability of LWR energy systems.

Wood, Thomas W.; Rothwell, Geoffrey

2012-01-10T23:59:59.000Z

111

Method for cleaning solution used in nuclear fuel reprocessing  

DOE Patents [OSTI]

Nuclear fuel processing solution consisting of tri-n-butyl phosphate and dodecane, with a complex of uranium, plutonium, or zirconium and with a solvent degradation product such as di-n-butyl phosphate therein, is contacted with an aqueous solution of a salt formed from hydrazine and either a dicarboxylic acid or a hydroxycarboxylic acid, thereby removing the aforesaid complex from the processing solution.

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

1980-12-17T23:59:59.000Z

112

Assessment of the effectiveness of mixed uranium-plutonium fuel in VVÉR  

Science Journals Connector (OSTI)

An assessment of the cost-effectiveness of burning mixed uranium-plutonium fuel in VVÉR reactors is made as a function of the price of natural uranium. It is shown that for the present price structure, based on t...

N. N. Ponomarev-Stepnoi; V. F. Tsibul’skii

2007-11-01T23:59:59.000Z

113

German Pebble Bed Research Reactor Highly Enriched Uranium (HEU...  

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

Potential Acceptance and Disposition of German Pebble Bed Research Reactor Highly Enriched Uranium (HEU) Fuel Environmental Assessment Maxcine Maxted, DOE-SR Used Nuclear Fuel...

114

A Neural Network Model for the Tomographic Analysis of Irradiated Nuclear Fuel Rods  

SciTech Connect (OSTI)

A tomographic method based on a multilayer feed-forward artificial neural network is proposed for the reconstruction of gamma-radioactive fission product distribution in irradiated nuclear fuel rods. The quality of the method is investigated as compared to a conventional technique on experimental results concerning a Canada deuterium uranium reactor (CANDU)-type fuel rod irradiated in a TRIGA reactor.

Craciunescu, Teddy [National Institute of Nuclear Physics and Engineering (Romania)

2004-04-15T23:59:59.000Z

115

The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication  

SciTech Connect (OSTI)

The friction bonding process combines desirable attributes of both friction stir welding and friction stir processing. The development of the process is spurred on by the need to fabricate thin, high density, reduced enrichment fuel plates for nuclear research reactors. The work seeks to convert research and test reactors currently operating on highly enriched uranium fuel to operate on low enriched uranium fuel without significant loss in reactor performance, safety characteristics, or significant increase in cost. In doing so, the threat of global nuclear material proliferation will be reduced. Feasibility studies performed on the process show that this is a viable option for mass production of plate-type nuclear fuel. Adapting the friction stir weld process for nuclear fuel fabrication has resulted in the development of several unique ideas and observations. Preliminary results of this adaptation and process model development are discussed.

D Burkes; P Medvedev; M Chapple; A Amritkar; P Wells; I Charit

2009-02-01T23:59:59.000Z

116

Engineering analysis of low enriched uranium fuel using improved zirconium hydride cross sections  

E-Print Network [OSTI]

for the change out of the existing high enriched uranium fuel to this high-burnup, low enriched uranium fuel design. The codes MCNP and Monteburns were utilized for the neutronic analysis while the code PARET was used to determine fuel and cladding temperatures...

Candalino, Robert Wilcox

2006-10-30T23:59:59.000Z

117

Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria  

SciTech Connect (OSTI)

The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

K. J. Allen; T. G. Apostolov; I. S. Dimitrov

2009-03-01T23:59:59.000Z

118

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

119

Nuclear Fuel Storage and Transportation Planning Project Overview...  

Office of Environmental Management (EM)

Fuel Storage and Transportation Planning Project Overview Nuclear Fuel Storage and Transportation Planning Project Overview Nuclear Fuel Storage and Transportation Planning Project...

120

Fuel Cycle Options for Optimized Recycling of Nuclear Fuel  

E-Print Network [OSTI]

The reduction of transuranic inventories of spent nuclear fuel depends upon the deployment of advanced fuels that can be loaded with recycled transuranics (TRU), and the availability of facilities to separate and reprocess ...

Aquien, A.

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Neutronics of accelerator-driven subcritical fission for burning transuranics in used nuclear fuel  

SciTech Connect (OSTI)

We report the development of a conceptual design for accelerator-driven subcritical fission in a molten salt core (ADSMS). ADSMS is capable of destroying all of the transuranics at the same rate and proportion as they are produced in a conventional nuclear power plant. The ADSMS core is fueled solely by transuranics extracted from used nuclear fuel and reduces its radiotoxicity by a factor 10,000. ADSMS offers a way to close the nuclear fuel cycle so that the full energy potential in the fertile fuels uranium and thorium can be recovered.

Sattarov, A.; Assadi, S.; Badgley, K.; Baty, A.; Comeaux, J.; Gerity, J.; Kellams, J.; Mcintyre, P.; Pogue, N.; Sooby, E.; Tsvetkov, P.; Rosaire, G. [Texas A and M University, College Station, TX 77845 (United States); Mann, T. [Argone National Laboratory, Argone, IL (United States)

2013-04-19T23:59:59.000Z

122

Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors` spent fuel  

SciTech Connect (OSTI)

One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE`s Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE`s efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE`s activities in taking back spent fuel.

NONE

1994-03-25T23:59:59.000Z

123

Greenhouse Gas Emissions from the Nuclear Fuel Cycle  

SciTech Connect (OSTI)

Since greenhouse gases are a global concern, rather than a local concern as are some kinds of effluents, one must compare the entire lifecycle of nuclear power to alternative technologies for generating electricity. A recent critical analysis by Sovacool (2008) gives a clearer picture. "It should be noted that nuclear power is not directly emitting greenhouse gas emissions, but rather that lifecycle emissions occur through plant construction, operation, uranium mining and milling, and plant decommissioning." "[N]uclear energy is in no way 'carbon free' or 'emissions free,' even though it is much better (from purely a carbon-equivalent emissions standpoint) than coal, oil, and natural gas electricity generators, but worse than renewable and small scale distributed generators" (Sovacool 2008). According to Sovacool, at an estimated 66 g CO2 equivalent per kilowatt-hour (gCO2e/kWh), nuclear power emits 15 times less CO2 per unit electricity generated than unscrubbed coal generation (at 1050 gCO2e/kWh), but 7 times more than the best renewable, wind (at 9 gCO2e/kWh). The U.S. Nuclear Regulatory Commission (2009) has long recognized CO2 emissions in its regulations concerning the environmental impact of the nuclear fuel cycle. In Table S-3 of 10 CFR 51.51(b), NRC lists a 1000-MW(electric) nuclear plant as releasing as much CO2 as a 45-MW(e) coal plant. A large share of the carbon emissions from the nuclear fuel cycle is due to the energy consumption to enrich uranium by the gaseous diffusion process. A switch to either gas centrifugation or laser isotope separation would dramatically reduce the carbon emissions from the nuclear fuel cycle.

Strom, Daniel J.

2010-03-01T23:59:59.000Z

124

Nuclear fuel cycle facility accident analysis handbook  

SciTech Connect (OSTI)

The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

NONE

1998-03-01T23:59:59.000Z

125

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

DOE Patents [OSTI]

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

Herrmann, Steven Douglas

2014-05-27T23:59:59.000Z

126

Utilization of Ceramic Uranium Fuels in ARIES-RS Fusion Reactor  

Science Journals Connector (OSTI)

This study presents the neutronic performance of the ARIES-RS fusion reactor design using different natural ceramic uranium fuels,...2, UN or U3Si2..., dispersed in graphite matrix. These fissionable fuels insert...

Mustafa Übeyli

2004-03-01T23:59:59.000Z

127

Spent Nuclear Fuel Self-Induced XRF to Predict Pu to U Content  

E-Print Network [OSTI]

area of interest, would improve input accountability and shipper/receiver differences. XRF measurements were made on individual PWR fuel rods with varying fuel ages and final burn-ups at Oak Ridge National Laboratory (ORNL) in July 2008 and January... Committee NRF Nuclear Resonance Fluorescence viii ORNL Oak Ridge National Laboratory Pu Plutonium PUREX Plutonium and Uranium Recovery by Extraction PWR Pressurized Water Reactor RPP Reprocessing Plant SNM Special Nuclear Material...

Stafford, Alissa Sarah

2010-10-12T23:59:59.000Z

128

Uranium Marketing Annual Report  

Gasoline and Diesel Fuel Update (EIA)

4. Uranium sellers to owners and operators of U.S. civilian nuclear power reactors, 2010-2012 2010 2011 2012 4. Uranium sellers to owners and operators of U.S. civilian nuclear power reactors, 2010-2012 2010 2011 2012 American Fuel Resources, LLC Advance Uranium Asset Management Ltd. (was Uranium Asset Management) Advance Uranium Asset Management Ltd. (was Uranium Asset Management) AREVA NC, Inc. (was COGEMA, Inc.) American Fuel Resources, LLC American Fuel Resources, LLC BHP Billiton Olympic Dam Corporation Pty Ltd AREVA NC, Inc. AREVA NC, Inc. CAMECO BHP Billiton Olympic Dam Corporation Pty Ltd BHP Billiton Olympic Dam Corporation Pty Ltd ConverDyn CAMECO CAMECO Denison Mines Corp. ConverDyn ConverDyn Energy Resources of Australia Ltd. Denison Mines Corp. Energy Fuels Resources Energy USA, Inc. Effective Energy N.V. Energy Resources of Australia Ltd.

129

Nuclear core and fuel assemblies  

DOE Patents [OSTI]

A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

Downs, Robert E. (Monroeville, PA)

1981-01-01T23:59:59.000Z

130

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides  

DOE Patents [OSTI]

Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Lloyd, M.H.

1981-01-09T23:59:59.000Z

131

Fuel cycle analysis of once-through nuclear systems.  

SciTech Connect (OSTI)

Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium (LEU) fuels. Examples of systems in this class include the small modular reactors being considered internationally; e.g. 4S [Tsuboi 2009], Hyperion Power Module [Deal 2010], ARC-100 [Wade 2010], and SSTAR [Smith 2008]. (2) Systems for Resource Utilization - In recent years, interest has developed in the use of advanced nuclear designs for the effective utilization of fuel resources. Systems under this class have generally utilized the breed and burn concept in which fissile material is bred and used in situ in the reactor core. Due to the favorable breeding that is possible with fast neutrons, these systems have tended to be fast spectrum systems. In the once-through concepts (as opposed to the traditional multirecycle approach typically considered for fast reactors), an ignition (or starter) zone contains driver fuel which is fissile material. This zone is designed to last a long time period to allow the breeding of sufficient fissile material in the adjoining blanket zone. The blanket zone is initially made of fertile depleted uranium fuel. This zone could also be made of fertile thorium fuel or recovered uranium from fuel reprocessing or natural uranium. However, given the bulk of depleted uranium and the potentially large inventory of recovered uranium, it is unlikely that the use of thorium is required in the near term in the U.S. Following the breeding of plutonium or fissile U-233 in the blanket, this zone or assembly then carries a larger fraction of the power generation in the reactor. These systems tend to also have a long cycle length (or core life) and they could be with or without fuel shuffling. When fuel is shuffled, the incoming fuel is generally depleted uranium (or thorium) fuel. In any case, fuel is burned once and then discharged. Examples of systems in this class include the CANDLE concept [Sekimoto 2001], the traveling wave reactor (TWR) concept of TerraPower [Ellis 2010], the ultra-long life fast reactor (ULFR) by ANL [Kim 2010], and the BNL fast mixed spectrum reactor (FMSR) concept [Fisher 1979]. (3) Thermal systems for resource extensio

Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

2010-08-10T23:59:59.000Z

132

Laboratory Directed Research and Development (LDRD) on Mono-uranium Nitride Fuel Development for SSTAR and Space Applications  

SciTech Connect (OSTI)

The US National Energy Policy of 2001 advocated the development of advanced fuel and fuel cycle technologies that are cleaner, more efficient, less waste-intensive, and more proliferation resistant. The need for advanced fuel development is emphasized in on-going DOE-supported programs, e.g., Global Nuclear Energy Initiative (GNEI), Advanced Fuel Cycle Initiative (AFCI), and GEN-IV Technology Development. The Directorates of Energy & Environment (E&E) and Chemistry & Material Sciences (C&MS) at Lawrence Livermore National Laboratory (LLNL) are interested in advanced fuel research and manufacturing using its multi-disciplinary capability and facilities to support a design concept of a small, secure, transportable, and autonomous reactor (SSTAR). The E&E and C&MS Directorates co-sponsored this Laboratory Directed Research & Development (LDRD) Project on Mono-Uranium Nitride Fuel Development for SSTAR and Space Applications. In fact, three out of the six GEN-IV reactor concepts consider using the nitride-based fuel, as shown in Table 1. SSTAR is a liquid-metal cooled, fast reactor. It uses nitride fuel in a sealed reactor vessel that could be shipped to the user and returned to the supplier having never been opened in its long operating lifetime. This sealed reactor concept envisions no fuel refueling nor on-site storage of spent fuel, and as a result, can greatly enhance proliferation resistance. However, the requirement for a sealed, long-life core imposes great challenges to research and development of the nitride fuel and its cladding. Cladding is an important interface between the fuel and coolant and a barrier to prevent fission gas release during normal and accidental conditions. In fabricating the nitride fuel rods and assemblies, the cladding material should be selected based on its the coolant-side corrosion properties, the chemical/physical interaction with the nitride fuel, as well as their thermal and neutronic properties. The US NASA space reactor, the SP-100 was designed to use mono-uranium nitride fuel. Although the SP-100 reactor was not commissioned, tens of thousand of nitride fuel pellets were manufactured and lots of them, cladded in Nb-1-Zr had been irradiated in fast test reactors (FFTF and EBR-II) with good irradiation results. The Russian Naval submarines also use nitride fuel with stainless steel cladding (HT-9) in Pb-Bi coolant. Although the operating experience of the Russian submarine is not readily available, such combination of fuel, cladding and coolant has been proposed for a commercial-size liquid-metal cooled fast reactor (BREST-300). Uranium mono-nitride fuel is studied in this LDRD Project due to its favorable properties such as its high actinide density and high thermal conductivity. The thermal conductivity of mono-nitride is 10 times higher than that of oxide (23 W/m-K for UN vs. 2.3 W/m-K for UO{sub 2} at 1000 K) and its melting temperature is much higher than that of metal fuel (2630 C for UN vs. 1132 C for U metal). It also has relatively high actinide density, (13.51 gU/cm{sup 3} in UN vs. 9.66 gU/cm{sup 3} in UO{sub 2}) which is essential for a compact reactor core design. The objective of this LDRD Project is to: (1) Establish a manufacturing capability for uranium-based ceramic nuclear fuel, (2) Develop a computational capability to analyze nuclear fuel performance, (3) Develop a modified UN-based fuel that can support a compact long-life reactor core, and (4) Collaborate with the Nuclear Engineering Department of UC Berkeley on nitride fuel reprocessing and disposal in a geologic repository.

Choi, J; Ebbinghaus, B; Meiers, T; Ahn, J

2006-02-09T23:59:59.000Z

133

Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process  

SciTech Connect (OSTI)

The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount of the resulting MOX. The study considered two sub-cases within each of the two fuel cycles in which the uranium and plutonium from the first generation of MOX spent fuel (i) would not be recycled to produce a second generation of MOX for use in LWRs or (ii) would be recycled to produce a second generation of MOX fuel for use in LWRs. The study also investigated the effects of recycling MOX spent fuel multiple times in LWRs. The study assumed that both fuel cycles would store and then reprocess spent MOX fuel that is not recycled to produce a next generation of LWR MOX fuel and would use the recovered products to produce FR fuel. The study further assumed that FRs would begin to be brought on-line in 2043, eleven years after recycle begins in LWRs, when products from 5-year cooled spent MOX fuel would be available. Fuel for the FRs would be made using the uranium, plutonium, and minor actinides recovered from MOX. For the cases where LWR fuel was assumed to be recycled one time, the 1st generation of MOX spent fuel was used to provide nuclear materials for production of FR fuel. For the cases where the LWR fuel was assumed to be recycled two times, the 2nd generation of MOX spent fuel was used to provide nuclear materials for production of FR fuel. The number of FRs in operation was assumed to increase in successive years until the rate that actinides were recovered from permanently discharged spent MOX fuel equaled the rate the actinides were consumed by the operating fleet of FRs. To compare the two fuel cycles, the study analyzed recycle of nuclear fuel in LWRs and FRs and determined the radiological characteristics of irradiated nuclear fuel, nuclear waste products, and recycle nuclear fuels. It also developed a model to simulate the flows of nuclear materials that could occur in the two advanced nuclear fuel cycles over 81 years beginning in 2020 and ending in 2100. Simulations projected the flows of uranium, plutonium, and minor actinides as these nuclear fuel materials were produced and consumed in a fleet of 100 1,000 MWe LWRs and in FRs. The model als

E. R. Johnson; R. E. Best

2009-12-28T23:59:59.000Z

134

Examining 239Pu and 240Pu Nuclear Resonance Fluorescence Measurements on Spent Fuel for Nuclear Safeguards  

E-Print Network [OSTI]

on Spent Fuel for Nuclear Safeguards Brian J. Quiter, ?resonances on nuclear safeguards measurements will be

Quiter, Brian

2013-01-01T23:59:59.000Z

135

Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating high-enriched uranium fuel core. These studies indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2009-11-01T23:59:59.000Z

136

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

137

Spent Nuclear Fuel (SNF) Project Execution Plan  

SciTech Connect (OSTI)

The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

LEROY, P.G.

2000-11-03T23:59:59.000Z

138

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

139

E-Print Network 3.0 - advanced lwr fuel Sample Search Results  

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

3 THE CASE FOR FUSION-FISSION HYBRIDS ENABLING SUSTAINABLE NUCLEAR POWER Summary: uranium energy content recovered in present LWR "once-through" fuel cycles (uranium would be...

140

TEPP - Spent Nuclear Fuel | Department of Energy  

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

- Spent Nuclear Fuel - Spent Nuclear Fuel TEPP - Spent Nuclear Fuel This scenario provides the planning instructions, guidance, and evaluation forms necessary to conduct an exercise involving a highway shipment of spent nuclear fuel. This exercise manual is one in a series of five scenarios developed by the Department of Energy Transportation Emergency Preparedness Program. Responding agencies may include several or more of the following: local municipal and county fire, police, sheriff, and Emergency Medical Services (EMS) personnel; state, local, and federal emergency response teams; emergency response contractors;and other emergency response resources that could potentially be provided by the carrier and the originating facility (shipper). Spent Nuclear Fuel.docx More Documents & Publications

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Interim report spent nuclear fuel retrieval system fuel handling development testing  

SciTech Connect (OSTI)

Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

1997-06-01T23:59:59.000Z

142

Test protocol for aluminum based spent nuclear fuel  

SciTech Connect (OSTI)

Aluminum based spent nuclear fuel (Al-SNF) will be treated and then stored in a geological repository. A series of corrosion tests, including coupon, galvanic, and electrochemical, were performed to measure the effects of degradation on U-Al alloys, which simulated various Al-SNF forms. The test solutions were variants of a J-13 well water chemistry. As part of the degradation characteristics, the dissolution of uranium, which is the primary radionuclide, was measured. The degradation and dissolution were changed by the type of water chemistry and the temperature. The alloy composition and fabrication were not significant variables.

Mickalonis, J.I.; Wiersma, B.J.

1999-07-01T23:59:59.000Z

143

Mississippi Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,251",8.0,"9,643",17.7 "Coal","2,526",16.1,"13,629",25.0 "Natural Gas","11,640",74.2,"29,619",54.4 "Other1",4,"*",10,"*" "Other Renewable1",235,1.5,"1,504",2.8 "Petroleum",35,0.2,81,0.1 "Total","15,691",100.0,"54,487",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

144

Iowa Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",601,4.1,"4,451",7.7 "Coal","6,956",47.7,"41,283",71.8 "Hydro and Pumped Storage",144,1.0,948,1.6 "Natural Gas","2,299",15.8,"1,312",2.3 "Other Renewable1","3,584",24.6,"9,360",16.3 "Petroleum","1,007",6.9,154,0.3 "Total","14,592",100.0,"57,509",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

145

Vermont Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",620,55.0,"4,782",72.2 "Hydro and Pumped Storage",324,28.7,"1,347",20.3 "Natural Gas","-","-",4,0.1 "Other Renewable1",84,7.5,482,7.3 "Petroleum",100,8.9,5,0.1 "Total","1,128",100.0,"6,620",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

146

Ohio Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,134",6.5,"15,805",11.0 "Coal","21,360",64.6,"117,828",82.1 "Hydro and Pumped Storage",101,0.3,429,0.3 "Natural Gas","8,203",24.8,"7,128",5.0 "Other1",123,0.4,266,0.2 "Other Renewable1",130,0.4,700,0.5 "Petroleum","1,019",3.1,"1,442",1.0 "Total","33,071",100.0,"143,598",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

147

Maryland Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,705",13.6,"13,994",32.1 "Coal","4,886",39.0,"23,668",54.3 "Hydro and Pumped Storage",590,4.7,"1,667",3.8 "Natural Gas","2,041",16.3,"2,897",6.6 "Other1",152,1.2,485,1.1 "Other Renewable1",209,1.7,574,1.3 "Petroleum","2,933",23.4,322,0.7 "Total","12,516",100.0,"43,607",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

148

Kansas Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,160",9.2,"9,556",19.9 "Coal","5,179",41.3,"32,505",67.8 "Hydro and Pumped Storage",3,"*",13,"*" "Natural Gas","4,573",36.5,"2,287",4.8 "Other Renewable1","1,079",8.6,"3,459",7.2 "Petroleum",550,4.4,103,0.2 "Total","12,543",100.0,"47,924",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

149

Connecticut Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,103",25.4,"16,750",50.2 "Coal",564,6.8,"2,604",7.8 "Hydro and Pumped Storage",151,1.8,400,1.2 "Natural Gas","2,292",27.7,"11,716",35.1 "Other1",27,0.3,730,2.2 "Other Renewable1",159,1.9,740,2.2 "Petroleum","2,989",36.1,409,1.2 "Total","8,284",100.0,"33,350",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

150

Nebraska Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,245",15.8,"11,054",30.2 "Coal","3,932",50.0,"23,363",63.8 "Hydro and Pumped Storage",278,3.5,"1,314",3.6 "Natural Gas","1,849",23.5,375,1.0 "Other Renewable1",165,2.1,493,1.3 "Petroleum",387,4.9,31,0.1 "Total","7,857",100.0,"36,630",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

151

Fuel cycle modelling of open cycle thorium-fuelled nuclear energy systems  

Science Journals Connector (OSTI)

Abstract In this study, we have sought to determine the advantages, disadvantages, and viability of open cycle thorium–uranium-fuelled (Th–U-fuelled) nuclear energy systems. This has been done by assessing three such systems, each of which requires uranium enriched to ?20% 235U, in comparison to a reference uranium-fuelled (U-fuelled) system over various performance indicators, spanning material flows, waste composition, economics, and proliferation resistance. The values of these indicators were determined using the UK National Nuclear Laboratory’s fuel cycle modelling code ORION. This code required the results of lattice-physics calculations to model the neutronics of each nuclear energy system, and these were obtained using various nuclear reactor physics codes and burn-up routines. In summary, all three Th–U-fuelled nuclear energy systems required more separative work capacity than the equivalent benchmark U-fuelled system, with larger levelised fuel cycle costs and larger levelised cost of electricity. Although a reduction of ?6% in the required uranium ore per kWh was seen for one of the Th–U-fuelled systems compared to the reference U-fuelled system, the other two Th–U-fuelled systems required more uranium ore per kWh than the reference. Negligible advantages and disadvantages were observed for the amount and the properties of the spent nuclear fuel (SNF) generated by the systems considered. Two of the Th–U-fuelled systems showed some benefit in terms of proliferation resistance of the SNF generated. Overall, it appears that there is little merit in incorporating thorium into nuclear energy systems operating with open nuclear fuel cycles.

S.F. Ashley; B.A. Lindley; G.T. Parks; W.J. Nuttall; R. Gregg; K.W. Hesketh; U. Kannan; P.D. Krishnani; B. Singh; A. Thakur; M. Cowper; A. Talamo

2014-01-01T23:59:59.000Z

152

Global Nuclear Energy Partnership Fact Sheet - Establish Reliable Fuel  

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

Global Nuclear Energy Partnership Fact Sheet - Establish Reliable Global Nuclear Energy Partnership Fact Sheet - Establish Reliable Fuel Services Global Nuclear Energy Partnership Fact Sheet - Establish Reliable Fuel Services GNEP would build and strengthen a reliable international fuel services consortium under which "fuel supplier nations" would choose to operate both nuclear power plants and fuel production and handling facilities, providing reliable fuel services to "user nations" that choose to only operate nuclear power plants. This international consortium is a critical component of the GNEP initiative to build an improved, more proliferation-resistant nuclear fuel cycle that recycles used fuel, while Global Nuclear Energy Partnership Fact Sheet - Establish Reliable Fuel Services More Documents & Publications

153

Nuclear Fuels Storage and Transportation Planning Project (NFST...  

Office of Environmental Management (EM)

Nuclear Fuels Storage and Transportation Planning Project (NFST) Program Status Nuclear Fuels Storage and Transportation Planning Project (NFST) Program Status Presentation made by...

154

2008 DOE Spent Nuclear Fuel and High Level Waste Inventory  

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

Management >> National Spent Nuclear Fuel INL Logo Search 2008 DOE Spent Nuclear Fuel and High Level Waste Inventory Content Goes Here Skip Navigation Links Home Newsroom About INL...

155

Department of Energy Awards $15 Million for Nuclear Fuel Cycle...  

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

Million for Nuclear Fuel Cycle Technology Research and Development Department of Energy Awards 15 Million for Nuclear Fuel Cycle Technology Research and Development August 1,...

156

Used Nuclear Fuel Loading and Structural Performance Under Normal...  

Office of Environmental Management (EM)

Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan Used Nuclear Fuel Loading...

157

Fuel handling apparatus for a nuclear reactor  

DOE Patents [OSTI]

Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

Hawke, Basil C. (Solana Beach, CA)

1987-01-01T23:59:59.000Z

158

Simulated nuclear reactor fuel assembly  

DOE Patents [OSTI]

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

Berta, V.T.

1993-04-06T23:59:59.000Z

159

Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel  

SciTech Connect (OSTI)

During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontally displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor metallic fuel will be shown to be of no consequence to reactor performance.

B.R. Westphal; D. Vaden; S.X. Li; G.L. Fredrickson; R.D. Mariani

2009-09-01T23:59:59.000Z

160

Fuel cycle options for optimized recycling of nuclear fuel  

E-Print Network [OSTI]

The accumulation of transuranic inventories in spent nuclear fuel depends on both deployment of advanced reactors that can be loaded with recycled transuranics (TRU), and on availability of the facilities that separate and ...

Aquien, Alexandre

2006-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Pyrochemical Treatment of Spent Nuclear Fuel  

SciTech Connect (OSTI)

Over the last 10 years, pyrochemical treatment of spent nuclear fuel has progressed from demonstration activities to engineering-scale production operations. As part of the Advanced Fuel Cycle Initiative within the U.S. Department of Energy’s Office of Nuclear Energy, Science and Technology, pyrochemical treatment operations are being performed as part of the treatment of fuel from the Experimental Breeder Reactor II at the Idaho National Laboratory. Integral to these treatment operations are research and development activities that are focused on scaling further the technology, developing and implementing process improvements, qualifying the resulting high-level waste forms, and demonstrating the overall pyrochemical fuel cycle.

K. M. Goff; K. L. Howden; G. M. Teske; T. A. Johnson

2005-10-01T23:59:59.000Z

162

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL  

SciTech Connect (OSTI)

This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recycling to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.

Matthew Bunn; Steve Fetter; John P. Holdren; Bob van der Zwaan

2003-07-01T23:59:59.000Z

163

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

DOE Patents [OSTI]

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

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

1999-03-23T23:59:59.000Z

164

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

DOE Patents [OSTI]

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

Crawford, Douglas C. (Idaho Falls, ID); Porter, Douglas L. (Idaho Falls, ID); Hayes, Steven L. (Idaho Falls, ID); Hill, Robert N. (Bolingbrook, IL)

1999-01-01T23:59:59.000Z

165

Overview of the nuclear fuel cycle  

SciTech Connect (OSTI)

The use of nuclear reactors to provide electrical energy has shown considerable growth since the first nuclear plant started commercial operation in the mid 1950s. Although the main purpose of this paper is to review the fuel cycle capabilities in the United States, the introduction is a brief review of the types of nuclear reactors in use and the world-wide nuclear capacity.

Leuze, R.E.

1982-01-01T23:59:59.000Z

166

Conditioning of spent nuclear fuel for permanent disposal  

SciTech Connect (OSTI)

A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or pyroprocessing, provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the US Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (> 99.9%) separation of transuranics. The resultant waste forms from the pyroprocess are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and that avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory.

Laidler, J.J. [Argonne National Lab., IL (United States). Chemical Technology Div.

1994-12-31T23:59:59.000Z

167

Conditioning of spent nuclear fuel for permanent disposal  

SciTech Connect (OSTI)

A compact, efficient method for conditioning spent nuclear fuel is under development This method, known as pyrochemical processing, or {open_quotes}pyroprocessing,{close_quotes} provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (99.9%) separation of transuranics. The resultant waste forms from the pyroprocess are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and preclude the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory.

Laidler, J.J.

1994-10-01T23:59:59.000Z

168

Fluorination of a depleted uranium-plutonium-nitride fuel with elemental fluorine  

Science Journals Connector (OSTI)

A physical and a mathematical model have been developed to describe the physicochemical process of torch fluorination of an uranium-plutonium-nitride fuel. An algorithm for calculating the velocity, temperatur...

V. A. Karelin; V. N. Brendakov; M. V. Popadeikin

169

Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities, Sections 15-19  

SciTech Connect (OSTI)

Information is presented under the following section headings: fuel reprocessing; spent fuel and high-level and transuranic waste storage; spent fuel and high-level and transuranic waste disposal; low-level and intermediate-level waste disposal; and, transportation of radioactive materials in the nuclear fuel cycle. In each of the first three sections a description is given on the mainline process, effluent processing and waste management systems, plant layout, and alternative process schemes. Safety information and a summary are also included in each. The section on transport of radioactive materials includes information on the transportation of uranium ore, uranium ore concentrate, UF/sub 6/, PuO/sub 2/ powder, unirradiated uranium and mixed-oxide fuel assemblies, spent fuel, solidified high-level waste, contact-handled transuranic waste, remote-handled transuranic waste, and low and intermediate level nontransuranic waste. A glossary is included. (JGB)

Schneider, K.J.

1982-09-01T23:59:59.000Z

170

Thermodynamics of high-temperature nuclear fuel  

Science Journals Connector (OSTI)

A method for performing a thermodynamic analysis of the high-temperature nuclear fuel using the ASTA computer program is substantiated. Calculations of the chemical composition and pressure of the gas phase of...

I. A. Belov; A. S. Ivanov

171

Chemistry of nuclear fuel reprocessing: Current status  

Science Journals Connector (OSTI)

Current status on the chemical aspects of nuclear fuel reprocessing is presented with special emphasis on the Purex process which continues to be the process of choice for the last four decades. Better deconta...

D. D. Sood; S. K. Patil

1996-03-01T23:59:59.000Z

172

Transuranium Elements in the Nuclear Fuel Cycle  

Science Journals Connector (OSTI)

Transuranium elements, neptunium, plutonium, americium, and curium, are formed via neutron capture processes of actinides, and are mainly by-products of fuel irradiation during the operation of a nuclear react...

Thomas Fanghänel; Jean-Paul Glatz; Rudy J. M. Konings…

2010-01-01T23:59:59.000Z

173

Fuel availability in nuclear power.  

E-Print Network [OSTI]

?? Nuclear power is in focus of attention due to several factors these days and the expression “nuclear renaissance” is getting well known. However, concerned… (more)

Söderlund, Karl

2009-01-01T23:59:59.000Z

174

Plutonium recovery from spent reactor fuel by uranium displacement  

DOE Patents [OSTI]

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

Ackerman, J.P.

1992-03-17T23:59:59.000Z

175

Fresh nuclear fuel measurements at Ukrainian nuclear power plants  

SciTech Connect (OSTI)

In 2005, the Provisions on Nuclear Material Measurement System was enacted in Ukraine as an important regulatory driver to support international obligations in nuclear safeguards and nonproliferation. It defines key provisions and requirements for material measurement and measurement control programs to ensure the quality and reliability of measurement data within the framework of the State MC&A System. Implementing the Provisions requires establishing a number of measurement techniques for both fresh and spent nuclear fuel for various types of Ukrainian reactors. Our first efforts focused on measurements of fresh nuclear fuel from a WWR-1000 power reactor.

Kuzminski, Jozef [Los Alamos National Laboratory; Ewing, Tom [ANL; Dickman, Debbie [PNNL; Gavrilyuk, Victor [UKRAINE; Drapey, Sergey [UKRAINE; Kirischuk, Vladimir [UKRAINE; Strilchuk, Nikolay [UKRAINE

2009-01-01T23:59:59.000Z

176

The chemical-induced genotoxicity of depleted uranium.  

E-Print Network [OSTI]

?? Uranium has been mined for many years and used for fuel for nuclear reactors and materials for atomic weapons, ammunition, and armor. While the… (more)

Yellowhair, Monica

2011-01-01T23:59:59.000Z

177

Determination of Uranium Metal Concentration in Irradiated Fuel Storage Basin Sludge Using Selective Dissolution  

SciTech Connect (OSTI)

Uranium metal corroding in water-saturated sludges now held in the US Department of Energy Hanford Site K West irradiated fuel storage basin can create hazardous hydrogen atmospheres during handling, immobilization, or subsequent transport and storage. Knowledge of uranium metal concentration in sludge thus is essential to safe sludge management and process design, requiring an expeditious routine analytical method to detect uranium metal concentrations as low as 0.03 wt% in sludge even in the presence of 30 wt% or higher total uranium concentrations.

Delegard, Calvin H.; Sinkov, Sergey I.; Chenault, Jeffrey W.; Schmidt, Andrew J.; Welsh, Terri L.; Pool, Karl N.

2014-03-01T23:59:59.000Z

178

Nuclear Fuel Cycle and Waste Management Technologies - Nuclear Engineering  

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

Nuclear Fuel Cycle and Nuclear Fuel Cycle and Waste Management Technologies Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Fuel Cycle and Waste Management Technologies Overview Bookmark and Share Much of the NE Division's research is directed toward developing software and performing analyses, system engineering design, and experiments to support the demonstration and optimization of the electrometallurgical

179

Idaho Site Completes Demolition of Cold War-era Nuclear Fuel Reprocessing  

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

Completes Demolition of Cold War-era Nuclear Fuel Completes Demolition of Cold War-era Nuclear Fuel Reprocessing Facility Idaho Site Completes Demolition of Cold War-era Nuclear Fuel Reprocessing Facility December 22, 2011 - 11:12am Addthis Media Contact Erik Simpson (208) 360-0426 A gravel mound, larger than half a city block and several feet thick, is the only visible feature that remains at the site of a Cold War-era spent nuclear fuel reprocessing facility at the U.S. Department of Energy's Idaho site. About $44 million in American Recovery and Reinvestment Act funds helped Idaho Cleanup Project crews accelerate the demolition of the facility that during its 40 years of operation recovered more than $1 billion worth of uranium. "The ability to retain our highly skilled workforce was a huge contributor to the success of this project," said Idaho Cleanup Project

180

Fuel cycle cost uncertainty from nuclear fuel cycle comparison  

SciTech Connect (OSTI)

This paper examined the uncertainty in fuel cycle cost (FCC) calculation by considering both model and parameter uncertainty. Four different fuel cycle options were compared in the analysis including the once-through cycle (OT), the DUPIC cycle, the MOX cycle and a closed fuel cycle with fast reactors (FR). The model uncertainty was addressed by using three different FCC modeling approaches with and without the time value of money consideration. The relative ratios of FCC in comparison to OT did not change much by using different modeling approaches. This observation was consistent with the results of the sensitivity study for the discount rate. Two different sets of data with uncertainty range of unit costs were used to address the parameter uncertainty of the FCC calculation. The sensitivity study showed that the dominating contributor to the total variance of FCC is the uranium price. In general, the FCC of OT was found to be the lowest followed by FR, MOX, and DUPIC. But depending on the uranium price, the FR cycle was found to have lower FCC over OT. The reprocessing cost was also found to have a major impact on FCC.

Li, J.; McNelis, D. [Institute for the Environment, University of North Carolina, Chapel Hill (United States); Yim, M.S. [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (Korea, Republic of)

2013-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Nuclear power and the fuel cycle  

Science Journals Connector (OSTI)

Due to rising energy costs and climate concerns, nuclear power is once again being seriously considered as an energy source by several countries. This revival of nuclear power is closely linked with the choice of fuel cycles available, and the intentions of countries pursuing nuclear power are likely to be, correctly or incorrectly, judged by the choice of fuel cycle they make. The needs and constraints of the emerging nuclear powers may, however, be different from the expectations of a segment of the world community. If this potential growth in nuclear power is not to be stifled, it is imperative that a climate of mutual trust is developed respecting every country's right to develop peaceful uses of nuclear power without leading to an atmosphere of mistrust regarding the 'intentions' behind the pursuit of peaceful nuclear power. While it will be a near impossibility to completely decouple the peaceful uses of nuclear power from its more destructive applications, it is important that aspiring countries develop a clear and transparent process. Technology-supplier countries also need to develop and follow clear and consistent treaties and national policies, avoiding ad hoc country-specific arrangements. We review here the state of interest in nuclear power and current policies and discuss fuel cycle options that may pave the way for the future growth of nuclear power.

Rizwan-uddin

2010-01-01T23:59:59.000Z

182

Measures of the Environmental Footprint of the Front End of the Nuclear Fuel Cycle  

SciTech Connect (OSTI)

Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle have focused primarily on energy consumption and CO2 emissions. Results have varied widely. Section 2 of this report provides a summary of historical estimates. This study revises existing empirical correlations and their underlying assumptions to fit to a more complete set of existing data. This study also addresses land transformation, water withdrawals, and occupational and public health impacts associated with the processes of the front end of the once-through nuclear fuel cycle. These processes include uranium mining, milling, refining, conversion, enrichment, and fuel fabrication. Metrics are developed to allow environmental impacts to be summed across the full set of front end processes, including transportation and disposition of the resulting depleted uranium.

Brett Carlsen; Emily Tavrides; Erich Schneider

2010-08-01T23:59:59.000Z

183

Nuclear Fuels & Materials Spotlight Volume 4  

SciTech Connect (OSTI)

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

2014-04-01T23:59:59.000Z

184

Uranium Mining Life-Cycle Energy Cost vs. Uranium Resources  

Science Journals Connector (OSTI)

The long-term viability of nuclear energy systems depends on the availability of uranium and on the question, whether the overall energy balance of the fuel cycle is positive, taking into account the full life-cy...

W. Eberhard Falck

2012-01-01T23:59:59.000Z

185

Development of monolithic nuclear fuels for RERTR by hot isostatic pressing  

SciTech Connect (OSTI)

The RERTR Program (Reduced Enrichment for Research and Test Reactors) is developing advanced nuclear fuels for high power test reactors. Monolithic fuel design provides a higher uranium loading than that of the traditional dispersion fuel design. In order to bond monolithic fuel meat to aluminum cladding, several bonding methods such as roll bonding, friction stir bonding and hot isostatic pressing, have been explored. Hot isostatic pressing is a promising process for low cost, batch fabrication of monolithic RERTR fuel plates. The progress on the development of this process at the Idaho National Laboratory will be presented. Due to the relatively high processing temperature used, the reaction between fuel meat and aluminum cladding to form brittle intermetallic phases may be a concern. The effect of processing temperature and time on the fuel/cladding reaction will be addressed. The influence of chemical composition on the reaction will also be discussed. (author)

Jue, J.-F.; Park, Blair; Chapple, Michael; Moore, Glenn; Keiser, Dennis [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

2008-07-15T23:59:59.000Z

186

Safety Aspects of Wet Storage of Spent Nuclear Fuel, OAS-L-13-11  

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

Safety Aspects of Wet Storage of Safety Aspects of Wet Storage of Spent Nuclear Fuel OAS-L-13-11 July 2013 Department of Energy Washington, DC 20585 July 10, 2013 MEMORANDUM FOR THE SENIOR ADVISOR FOR ENVIRONMENTAL MANAGEMENT FROM: Daniel M. Weeber Assistant Inspector General for Audits and Administration Office of Inspector General SUBJECT: INFORMATION: Audit Report on "Safety Aspects of Wet Storage of Spent Nuclear Fuel" BACKGROUND The Department of Energy (Department) is responsible for managing and storing spent nuclear fuel (SNF) generated by weapons and research programs and recovered through nonproliferation programs. The SNF consists of irradiated reactor fuel and cut up assemblies containing uranium, thorium and/or plutonium. The Department stores 34 metric tons of heavy metal SNF primarily

187

International Nuclear Fuel Cycle Fact Book  

SciTech Connect (OSTI)

As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need exists costs for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book has been compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NMEA activities reports; and proceedings of conferences and workshops. The data listed typically do not reflect any single source but frequently represent a consolidation/combination of information.

Leigh, I.W.

1992-05-01T23:59:59.000Z

188

Safeguarding and Protecting the Nuclear Fuel Cycle  

SciTech Connect (OSTI)

International safeguards as applied by the International Atomic Energy Agency (IAEA) are a vital cornerstone of the global nuclear nonproliferation regime - they protect against the peaceful nuclear fuel cycle becoming the undetected vehicle for nuclear weapons proliferation by States. Likewise, domestic safeguards and nuclear security are essential to combating theft, sabotage, and nuclear terrorism by non-State actors. While current approaches to safeguarding and protecting the nuclear fuel cycle have been very successful, there is significant, active interest to further improve the efficiency and effectiveness of safeguards and security, particularly in light of the anticipated growth of nuclear energy and the increase in the global threat environment. This article will address two recent developments called Safeguards-by-Design and Security-by-Design, which are receiving increasing broad international attention and support. Expected benefits include facilities that are inherently more economical to effectively safeguard and protect. However, the technical measures of safeguards and security alone are not enough - they must continue to be broadly supported by dynamic and adaptive nonproliferation and security regimes. To this end, at the level of the global fuel cycle architecture, 'nonproliferation and security by design' remains a worthy objective that is also the subject of very active, international focus.

Trond Bjornard; Humberto Garcia; William Desmond; Scott Demuth

2010-11-01T23:59:59.000Z

189

Double-clad nuclear fuel safety rod  

DOE Patents [OSTI]

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

McCarthy, William H. (Los Altos, CA); Atcheson, Donald B. (Cupertino, CA); Vaidyanathan, Swaminathan (San Jose, CA)

1984-01-01T23:59:59.000Z

190

Resource intensities of the front end of the nuclear fuel cycle  

SciTech Connect (OSTI)

This paper presents resource intensities, including direct and embodied energy consumption, land and water use, associated with the processes comprising the front end of the nuclear fuel cycle. These processes include uranium extraction, conversion, enrichment, fuel fabrication and depleted uranium de-conversion. To the extent feasible, these impacts are calculated based on data reported by operating facilities, with preference given to more recent data based on current technologies and regulations. All impacts are normalized per GWh of electricity produced. Uranium extraction is seen to be the most resource intensive front end process. Combined, the energy consumed by all front end processes is equal to less than 1% of the electricity produced by the uranium in a nuclear reactor. Land transformation and water withdrawals are calculated at 8.07 m{sup 2} /GWh(e) and 1.37x10{sup 5} l/GWh(e), respectively. Both are dominated by the requirements of uranium extraction, which accounts for over 70% of land use and nearly 90% of water use.

Schneider, E.; Phathanapirom, U. [The University of Texas at Austin, 1 University Station C2200, Austin TX 78712 (United States); Eggert, R.; Collins, J. [Colorado School of Mines, 1500 Illinois St., Golden CO 80401 (United States)

2013-07-01T23:59:59.000Z

191

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

192

Nuclear fuel elements made from nanophase materials  

DOE Patents [OSTI]

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

Heubeck, N.B.

1998-09-08T23:59:59.000Z

193

Synthesis of uranium nitride and uranium carbide powder by carbothermic reduction  

SciTech Connect (OSTI)

Uranium nitride and uranium carbide are being considered as high burnup fuels in next generation nuclear reactors and accelerated driven systems for the transmutation of nuclear waste. The same characteristics that make nitrides and carbides candidates for these applications (i.e. favorable thermal properties, mutual solubility of nitrides, etc.), also make these compositions candidate fuels for space nuclear reactors. In this paper, we discuss the synthesis and characterization of depleted uranium nitride and carbide for a space nuclear reactor program. Importantly, this project emphasized that to synthesize high quality uranium nitride and carbide, it is necessary to understand the exact stoichiometry of the oxide feedstock. (authors)

Dunwoody, J.T.; Stanek, C.R.; McClellan, K.J.; Voit, S.L.; Volz, H.M. [Los Alamos National Laboratory, Los Alamos, New Mexico (United States); Hickman, R.R. [NASA Marshall Space Flight Center, Huntsville, Alabama (United States)

2007-07-01T23:59:59.000Z

194

Commercial nuclear fuel from U.S. and Russian surplus defense inventories: Materials, policies, and market effects  

SciTech Connect (OSTI)

Nuclear materials declared by the US and Russian governments as surplus to defense programs are being converted into fuel for commercial nuclear reactors. This report presents the results of an analysis estimating the market effects that would likely result from current plans to commercialize surplus defense inventories. The analysis focuses on two key issues: (1) the extent by which traditional sources of supply, such as production from uranium mines and enrichment plants, would be displaced by the commercialization of surplus defense inventories or, conversely, would be required in the event of disruptions to planned commercialization, and (2) the future price of uranium considering the potential availability of surplus defense inventories. Finally, the report provides an estimate of the savings in uranium procurement costs that could be realized by US nuclear power generating companies with access to competitively priced uranium supplied from surplus defense inventories.

NONE

1998-05-01T23:59:59.000Z

195

Cost Savings of Nuclear Power with Total Fuel Reprocessing  

SciTech Connect (OSTI)

The cost of fast reactor (FR) generated electricity with pyro-processing is estimated in this article. It compares favorably with other forms of energy and is shown to be less than that produced by light water reactors (LWR's). FR's use all the energy in natural uranium whereas LWR's utilize only 0.7% of it. Because of high radioactivity, pyro-processing is not open to weapon material diversion. This technology is ready now. Nuclear power has the same advantage as coal power in that it is not dependent upon a scarce foreign fuel and has the significant additional advantage of not contributing to global warming or air pollution. A jump start on new nuclear plants could rapidly allow electric furnaces to replace home heating oil furnaces and utilize high capacity batteries for hybrid automobiles: both would reduce US reliance on oil. If these were fast reactors fueled by reprocessed fuel, the spent fuel storage problem could also be solved. Costs are derived from assumptions on the LWR's and FR's five cost components: 1) Capital costs: LWR plants cost $106/MWe. FR's cost 25% more. Forty year amortization is used. 2) The annual O and M costs for both plants are 9% of the Capital Costs. 3) LWR fuel costs about 0.0035 $/kWh. Producing FR fuel from spent fuel by pyro-processing must be done in highly shielded hot cells which is costly. However, the five foot thick concrete walls have the advantage of prohibiting diversion. LWR spent fuel must be used as feedstock for the FR initial core load and first two reloads so this FR fuel costs more than LWR fuel. FR fuel costs much less for subsequent core reloads (< LWR fuel) if all spent fuel feedstock is from the fast reactor (i.e., Breeding Ratio =1). 4) Yucca Mountain storage of unprocessed LWR spent fuel is estimated as $360,000/MTHM. But this fuel can be processed to remove TRU for use as fast reactor fuel. The remaining fission products repository costs are only one fifth that of the original fuel. Storage of short half life fission products alone requires less storage time and long term integrity than LWR spent fuel (300 years storage versus 100,000 years.) 5) LWR decommissioning costs are estimated to be $0.3 x 10{sup 6}/MWe. The annual cost for a 40 year licensed plant would be 2.5 % of this or less if interest is taken into account. All plants will eventually have to replace those components which become radiation damaged. FR's should be designed to replace parts rather than decommission. The LWR costs are estimated to be 2.65 cents/kWh. FR costs are 2.99 cents/kWh for the first 7.5 years and 2.39 cents/kWh for the next 32.5 years. The average cost over forty years is 2.50 cents/kWh which is less than the LWR costs. These power costs are similar to coal power, are lower than gas, oil, and much lower than renewable power.(authors)

Solbrig, Charles W.; Benedict, Robert W. [Fuel Cycle Programs Division, Idaho National Laboratory, Idaho Falls, Idaho (United States)

2006-07-01T23:59:59.000Z

196

Actinides in metallic waste from electrometallurgical treatment of spent nuclear fuel.  

SciTech Connect (OSTI)

Argonne National Laboratory has developed a pyroprocessing-based technique for conditioning spent sodium-bonded nuclear-reactor fuel in preparation for long-term disposal. The technique produces a metallic waste form whose nominal composition is stainless steel with 15 wt.% Zr (SS-15Zr), up to {approx} 11 wt.% actinide elements (primarily uranium), and a few percent metallic fission products. Actual and simulated waste forms show similar eutectic microstructures with approximately equal proportions of iron solid solution phases and Fe-Zr intermetallics. This article reports on an analysis of simulated waste forms containing uranium, neptunium, and plutonium.

Janney, D. E.; Keiser, D. D., Jr.; Engineering Technology

2003-09-01T23:59:59.000Z

197

A review of uranium economics  

Science Journals Connector (OSTI)

The recent increase in the demand for power for commercial use, the challenges facing fossil fuel use and the prospective of cheap nuclear power motivate different countries to plan for the use of nuclear power. This paper reviews many aspects of uranium economics, which includes the advantages and disadvantages of nuclear power, comparisons with other sources of power, nuclear power production and requirements, the uranium market, uranium pricing, spot price and long-term price indicators, and the cost of building a nuclear power facility.

A.K. Mazher

2009-01-01T23:59:59.000Z

198

Michigan Nuclear Profile - All Fuels  

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

total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,947",13.2,"29,625",26.6 "Coal","11,531",38.7,"65,604",58.8 "Hydro and Pumped...

199

Minnesota Nuclear Profile - All Fuels  

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

total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,594",10.8,"13,478",25.1 "Coal","4,789",32.5,"28,083",52.3 "Hydro and Pumped...

200

Wisconsin Nuclear Profile - All Fuels  

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

total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,584",8.9,"13,281",20.7 "Coal","8,063",45.2,"40,169",62.5 "Hydro and Pumped...

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Washington Nuclear Profile - All Fuels  

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

total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,097",3.6,"9,241",8.9 "Coal","1,340",4.4,"8,527",8.2 "Hydro and Pumped...

202

Virginia Nuclear Profile - All Fuels  

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

total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,501",14.5,"26,572",36.4 "Coal","5,868",24.3,"25,459",34.9 "Hydro and Pumped...

203

A Characteristics-Based Approach to Radioactive Waste Classification in Advanced Nuclear Fuel Cycles  

E-Print Network [OSTI]

Nuclear   Fuel”,   Nuclear  Engineering  and  Technology,  in   Engineering  -­?  Nuclear  Engineering   and  the  in  Engineering  -­?  Nuclear  Engineering   and  the  

Djokic, Denia

2013-01-01T23:59:59.000Z

204

Nuclear Fuels Storage & Transportation Planning Project Documents |  

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

Fuel Cycle Technologies » Nuclear Fuels Storage & Fuel Cycle Technologies » Nuclear Fuels Storage & Transportation Planning Project » Nuclear Fuels Storage & Transportation Planning Project Documents Nuclear Fuels Storage & Transportation Planning Project Documents September 30, 2013 Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites In January 2013, the Department of Energy issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. February 22, 2013 Public Preferences Related to Consent-Based Siting of Radioactive Waste Management Facilities for Storage and Disposal This report provides findings from a set of social science studies

205

Seawater Enhances the Corrosion of Nuclear Fuel Rods | Photosynthetic...  

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

Seawater Enhances the Corrosion of Nuclear Fuel Rods April 19, 2012 Seawater Enhances the Corrosion of Nuclear Fuel Rods PARC Post Doc Anne-Marie Carey is featured in DOE Frontiers...

206

Strategy for the Management and Disposal of Used Nuclear Fuel...  

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

Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level...

207

Surrogate Spent Nuclear Fuel Vibration Integrity Investigation  

SciTech Connect (OSTI)

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Bevard, Bruce Balkcom [ORNL; Howard, Rob L [ORNL

2014-01-01T23:59:59.000Z

208

Spent Nuclear Fuel Alternative Technology Decision Analysis  

SciTech Connect (OSTI)

The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

Shedrow, C.B.

1999-11-29T23:59:59.000Z

209

US-Russian collaboration in MPC & A enhancements at the Elektrostal Uranium Fuel-Fabrication Plant  

SciTech Connect (OSTI)

Enhancement of the nuclear materials protection, control, and accounting of (MPC&A) at the Elektrostal Machine-Building Plant (ELEMASH) has proceeded in two phases. Initially, Elektrostal served as the model facility at which to test US/Russian collaboration and to demonstrate MPC&A technologies available for safeguards enhancements at Russian facilities. This phase addressed material control and accounting (MC&A) in the low-enriched uranium (LEU) fuel-fabrication processes and the physical protection (PP) of part of the (higher-enrichment) breeder-fuel process. The second phase, identified later in the broader US/Russian agreement for expanded MPC&A cooperation. includes implementation of appropriate MC&A and PP systems in the breeder-fuel fabrication processes. Within the past year, an automated physical protection system has been installed and demonstrated in building 274, and an automated MC&A system has been designed and is being installed and will be tested in the LEU process. Attention has now turned to assuring longterm sustainability for the first phase and beginning MPC&A upgrades for the second phase. Sustainability measures establish the infrastructure for operation, maintenance, and repair of the installed systems-with US support for the lifetime of the US/Russian Agreement, but evolving toward full Russian operation of the system over the long term. For phase 2, which will address higher enrichments, projects have been identified to characterize the facilities, design MPC&A systems, procure appropriate equipment, and install and test final systems. One goal in phase 2 will be to build on initial work to create shared, plant-wide MPC&A assets for operation, maintenance, and evaluation of all safeguards systems.

Smith, H.; Murray, W.; Whiteson, R. [and others

1997-11-01T23:59:59.000Z

210

E-Print Network 3.0 - americium plutonium uranium Sample Search...  

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

Beyond the Myth Summary: a powerful nuclear fuel group that would master everything from uranium mining to plutonium fuels, civil... in principle to the spent fuel sent to La...

211

Measures of the environmental footprint of the front end of the nuclear fuel cycle  

SciTech Connect (OSTI)

Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle (FEFC) have focused primarily on energy consumption and CO2 emissions. Results have varied widely. This work builds upon reports from operating facilities and other primary data sources to build a database of front end environmental impacts. This work also addresses land transformation and water withdrawals associated with the processes of the FEFC. These processes include uranium extraction, conversion, enrichment, fuel fabrication, depleted uranium disposition, and transportation. To allow summing the impacts across processes, all impacts were normalized per tonne of natural uranium mined as well as per MWh(e) of electricity produced, a more conventional unit for measuring environmental impacts that facilitates comparison with other studies. This conversion was based on mass balances and process efficiencies associated with the current once-through LWR fuel cycle. Total energy input is calculated at 8.7 x 10- 3 GJ(e)/MWh(e) of electricity and 5.9 x 10- 3 GJ(t)/MWh(e) of thermal energy. It is dominated by the energy required for uranium extraction, conversion to fluoride compound for subsequent enrichment, and enrichment. An estimate of the carbon footprint is made from the direct energy consumption at 1.7 kg CO2/MWh(e). Water use is likewise dominated by requirements of uranium extraction, totaling 154 L/MWh(e). Land use is calculated at 8 x 10- 3 m2/MWh(e), over 90% of which is due to uranium extraction. Quantified impacts are limited to those resulting from activities performed within the FEFC process facilities (i.e. within the plant gates). Energy embodied in material inputs such as process chemicals and fuel cladding is identified but not explicitly quantified in this study. Inclusion of indirect energy associated with embodied energy as well as construction and decommissioning of facilities could increase the FEFC energy intensity estimate by a factor of up to 2.

E. Schneider; B. Carlsen; E. Tavrides; C. van der Hoeven; U. Phathanapirom

2013-11-01T23:59:59.000Z

212

Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel  

SciTech Connect (OSTI)

The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

Not Available

1994-04-01T23:59:59.000Z

213

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing  

SciTech Connect (OSTI)

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

D.R. Jackson; G.R. Kiebel

1999-08-24T23:59:59.000Z

214

Alternative dispositioning methods for HEU spent nuclear fuel at the Savannah River Site  

SciTech Connect (OSTI)

The United States has a strong policy on prevention of the international spread of nuclear weapons. This policy was announced in Presidential Directive PDD-13 and summarized in a White House press release September 27, 1993. Two cornerstones of this policy are: seek to eliminate where possible the accumulation of stockpiles of highly- enriched uranium or plutonium; propose{hor_ellipsis}prohibiting the production of highly-enriched uranium (HEU) or plutonium for nuclear explosives purposes or outside international safeguards. The Department of Energy is currently struggling to devise techniques that safely and efficiently dispose of spent nuclear fuel (SNF) while satisfying national non-proliferation policies. SRS plans and proposals for disposing of their SNF are safe and cost effective, and fully satisfy non-proliferation objectives.

Krupa, J.F.; McKibben, J.M.; Parks, P.B.; DuPont, M.E.

1995-11-01T23:59:59.000Z

215

Evacuation and Shelter in Place Modeling for a Release of Uranium Hexafluoride.  

E-Print Network [OSTI]

?? Evacuation and sheltering behaviors were modeled for a hypothetical release of uranium hexafluoride (UF6) from Nuclear Fuel Services (NFS) in Erwin, Tennessee. NFS down-blends… (more)

Harris, Joseph B

2014-01-01T23:59:59.000Z

216

Ultraslow Wave Nuclear Burning of Uranium-Plutonium Fissile Medium on Epithermal Neutrons  

E-Print Network [OSTI]

For a fissile medium, originally consisting of uranium-238, the investigation of fulfillment of the wave burning criterion in a wide range of neutron energies is conducted for the first time, and a possibility of wave nuclear burning not only in the region of fast neutrons, but also for cold, epithermal and resonance ones is discovered for the first time. For the first time the results of the investigation of the Feoktistov criterion fulfillment for a fissile medium, originally consisting of uranium-238 dioxide with enrichments 4.38%, 2.00%, 1.00%, 0.71% and 0.50% with respect to uranium-235, in the region of neutron energies 0.015-10.0eV are presented. These results indicate a possibility of ultraslow wave neutron-nuclear burning mode realization in the uranium-plutonium media, originally (before the wave initiation by external neutron source) having enrichments with respect to uranium-235, corresponding to the subcritical state, in the regions of cold, thermal, epithermal and resonance neutrons. In order to validate the conclusions, based on the slow wave neutron-nuclear burning criterion fulfillment depending on the neutron energy, the numerical modeling of ultraslow wave neutron-nuclear burning of a natural uranium in the epithermal region of neutron energies (0.1-7.0eV) was conducted for the first time. The presented simulated results indicate the realization of the ultraslow wave neutron-nuclear burning of the natural uranium for the epithermal neutrons.

V. D. Rusov; V. A. Tarasov; M. V. Eingorn; S. A. Chernezhenko; A. A. Kakaev; V. M. Vashchenko; M. E. Beglaryan

2014-09-25T23:59:59.000Z

217

VISION - Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics  

SciTech Connect (OSTI)

The U.S. DOE Advanced Fuel Cycle Initiative’s (AFCI) fundamental objective is to provide technology options that - if implemented - would enable long-term growth of nuclear power while improving sustainability and energy security. The AFCI organization structure consists of four areas; Systems Analysis, Fuels, Separations and Transmutations. The Systems Analysis Working Group is tasked with bridging the program technical areas and providing the models, tools, and analyses required to assess the feasibility of design and deployment options and inform key decision makers. An integral part of the Systems Analysis tool set is the development of a system level model that can be used to examine the implications of the different mixes of reactors, implications of fuel reprocessing, impact of deployment technologies, as well as potential "exit" or "off ramp" approaches to phase out technologies, waste management issues and long-term repository needs. The Verifiable Fuel Cycle Simulation Model (VISION) is a computer-based simulation model that allows performing dynamic simulations of fuel cycles to quantify infrastructure requirements and identify key trade-offs between alternatives. It is based on the current AFCI system analysis tool "DYMOND-US" functionalities in addition to economics, isotopic decay, and other new functionalities. VISION is intended to serve as a broad systems analysis and study tool applicable to work conducted as part of the AFCI and Generation IV reactor development studies.

Steven J. Piet; A. M. Yacout; J. J. Jacobson; C. Laws; G. E. Matthern; D. E. Shropshire

2006-02-01T23:59:59.000Z

218

DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear  

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

to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile November 7, 2005 - 12:38pm Addthis Will Be Redirected to Naval Reactors, Down-blended or Used for Space Programs WASHINGTON, DC - Secretary of Energy Samuel W. Bodman today announced that the Department of Energy's (DOE) National Nuclear Security Administration (NNSA) will remove up to 200 metric tons (MT) of Highly Enriched Uranium (HEU), in the coming decades, from further use as fissile material in U.S. nuclear weapons and prepare this material for other uses. Secretary Bodman made this announcement while addressing the 2005 Carnegie International Nonproliferation Conference in Washington, DC.

219

Engineering of Deinococcus radiodurans R1 for Bioprecipitation of Uranium from Dilute Nuclear Waste  

Science Journals Connector (OSTI)

...February 15, 2007 ERRATUM ERRATUM Engineering of Deinococcus radiodurans R1 for Bioprecipitation of Uranium from Dilute Nuclear Waste Deepti Appukuttan Amara Sambasiva Rao Shree Kumar Apte Molecular Biology Division, Bhabha Atomic Research Centre...

Deepti Appukuttan; Amara Sambasiva Rao; Shree Kumar Apte

2007-02-15T23:59:59.000Z

220

Surface water transport and distribution of uranium in contaminated sediments near a nuclear weapons processing facility  

E-Print Network [OSTI]

The extent of remobilization of uranium from contaminated soils adjacent to a nuclear weapons processing facility during episodic rain events was investigated. In addition, information on the solid phase associations of U in floodplain and suspended...

Batson, Vicky Lynn

1994-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

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

Science Journals Connector (OSTI)

The production of enriched uranium used in nuclear weapons and fuel for ... power plants is accompanied by the formation of depleted uranium (DU), the amount of which annually ... DU mass is stored as environ-men...

N. P. Laverov; V. I. Velichkin; B. I. Omel’yanenko…

2010-08-01T23:59:59.000Z

222

Submersion Criticality Safety Analysis of Tungsten-Based Fuel for Nuclear Power and Propulsion Applications  

SciTech Connect (OSTI)

The Center for Space Nuclear Research (CSNR) is developing tungsten-encapsulated fuels for space nuclear applications. Aims to develop NTP fuels that are; Affordable Low impact on production and testing environment Producible on a large scale over suitable time period Higher-performance compared to previous graphite NTP fuel elements Space nuclear reactors remain subcritical before and during launch, and do not go critical until required by its mission. A properly designed reactor will remain subcritical in any launch abort scenario, where the reactor falls back to Earth and becomes submerged in terrestrial material. Submersion increases neutron reflection and thermalizes the neutrons, which typically increases the reactivity of the core. This effect is usually very significant for fast-spectrum reactors. This research provided a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor. Determine the submersion behavior of a reactor fueled by tungsten-based fuel. Considered fuel compositions with varying: Rhenium content (wt% rhenium in tungsten) Fuel loading fractions (UO2 vol%)

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

2014-07-01T23:59:59.000Z

223

Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel  

SciTech Connect (OSTI)

Performance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO{sub 2} matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron X-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO{sub 2} matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO{sub 2}{sup 2+}) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of {approx} 50 micrometers. We find evidence of a thin ({approx} 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO{sub 2}{sup +}/Np{sup 4+} and UO{sub 2}{sup 2+}/U{sup 4+} couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. (authors)

Fortner, Jeffrey A.; Kropf, A. Jeremy; Jerden, James L.; Cunnane, James C. [Chemical Engineering, Argonne National Laboratory, CMT/205, 9700 S. Cass Avenue, Argonne, IL, 60439 (United States)

2007-07-01T23:59:59.000Z

224

Texas Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,966",4.6,"41,335",10.0 "Coal","22,335",20.6,"150,173",36.5 "Hydro and Pumped Storage",689,0.6,"1,262",0.3 "Natural Gas","69,291",64.0,"186,882",45.4 "Other1",477,0.4,"3,630",0.9 "Other Renewable1","10,295",9.5,"27,705",6.7 "Petroleum",204,0.2,708,0.2 "Total","108,258",100.0,"411,695",100.0

225

Pennsylvania Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","9,540",20.9,"77,828",33.9 "Coal","18,481",40.6,"110,369",48.0 "Hydro and Pumped Storage","2,268",5.0,"1,624",0.7 "Natural Gas","9,415",20.7,"33,718",14.7 "Other1",100,0.2,"1,396",0.6 "Other Renewable1","1,237",2.7,"4,245",1.8 "Petroleum","4,534",9.9,571,0.2 "Total","45,575",100.0,"229,752",100.0

226

California Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,390",6.5,"32,201",15.8 "Coal",374,0.6,"2,100",1.0 "Hydro and Pumped Storage","13,954",20.7,"33,260",16.3 "Natural Gas","41,370",61.4,"107,522",52.7 "Other1",220,0.3,"2,534",1.2 "Other Renewable1","6,319",9.4,"25,450",12.5 "Petroleum",701,1.0,"1,059",0.5 "Total","67,328",100.0,"204,126",100.0

227

Arizona Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (nw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",3937,14.9,"31,200",27.9 "Coal","6,233",23.6,"43,644",39.1 "Hydro and Pumped Storage","2,937",11.1,"6,831",6.1 "Natural Gas","13,012",49.3,"29,676",26.6 "Other1","-","-",15,"*" "Other Renewable1",181,0.7,319,0.3 "Petroleum",93,0.4,66,0.1 "Total","26,392",100.0,"111,751",100.0

228

Louisiana Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,142",8.0,"18,639",18.1 "Coal","3,417",12.8,"23,924",23.3 "Hydro and Pumped Storage",192,0.7,"1,109",1.1 "Natural Gas","19,574",73.2,"51,344",49.9 "Other1",213,0.8,"2,120",2.1 "Other Renewable1",325,1.2,"2,468",2.4 "Petroleum",881,3.3,"3,281",3.2 "Total","26,744",100.0,"102,885",100.0

229

Illinois Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","11,441",25.9,"96,190",47.8 "Coal","15,551",35.2,"93,611",46.5 "Hydro and Pumped Storage",34,0.1,119,0.1 "Natural Gas","13,771",31.2,"5,724",2.8 "Other1",145,0.3,461,0.2 "Other Renewable1","2,078",4.7,"5,138",2.6 "Petroleum","1,106",2.5,110,0.1 "Total","44,127",100.0,"201,352",100.0

230

Alabama Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","5,043",15.6,"37,941",24.9 "Coal","11,441",35.3,"63,050",41.4 "Hydro and Pumped Storage","3,272",10.1,"8,704",5.7 "Natural Gas","11,936",36.8,"39,235",25.8 "Other1",100,0.3,643,0.4 "Other Renewable1",583,1.8,"2,377",1.6 "Petroleum",43,0.1,200,0.1 "Total","32,417",100.0,"152,151",100.0

231

Florida Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (nw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,924",6.6,"23,936",10.4 "Coal","9,975",16.9,"59,897",26.1 "Hydro and Pumped Storage",55,0.1,177,0.1 "Natural Gas","31,563",53.4,"128,634",56.1 "Other1",544,0.9,"2,842",1.2 "Other Renewable1","1,053",1.8,"4,487",2.0 "Petroleum","12,033",20.3,"9,122",4.0 "Total","59,147",100.0,"229,096",100.0

232

Arkansas Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,835",11.5,"15,023",24.6 "Coal","4,535",28.4,"28,152",46.2 "Hydro and Pumped Storage","1,369",8.6,"3,658",6.0 "Natural Gas","7,894",49.4,"12,469",20.4 "Other1","-","-",28,"*" "Other Renewable1",326,2.0,"1,624",2.7 "Petroleum",22,0.1,45,0.1 "Total","15,981",100.0,"61,000",100.0

233

Missouri Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,190",5.5,"8,996",9.7 "Coal","12,070",55.5,"75,047",81.3 "Hydro and Pumped Storage","1,221",5.6,"2,427",2.6 "Natural Gas","5,579",25.7,"4,690",5.1 "Other1","-","-",39,"*" "Other Renewable1",466,2.1,988,1.1 "Petroleum","1,212",5.6,126,0.1 "Total","21,739",100.0,"92,313",100.0

234

Massachusetts Nuclear Profile - All Fuels  

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

total electric power industry, smmer capacity and net generation, by energy source, 2010" total electric power industry, smmer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",685,5.0,"5,918",13.8 "Coal","1,669",12.2,"8,306",19.4 "Hydro and Pumped Storage","1,942",14.2,659,1.5 "Natural Gas","6,063",44.3,"25,582",59.8 "Other1",3,"*",771,1.8 "Other Renewable1",304,2.2,"1,274",3.0 "Petroleum","3,031",22.1,296,0.7 "Total","13,697",100.0,"42,805",100.0

235

Georgia Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,061",11.1,"33,512",24.4 "Coal","13,230",36.1,"73,298",53.3 "Hydro and Pumped Storage","3,851",10.5,"3,044",2.2 "Natural Gas","12,668",34.6,"23,884",17.4 "Other1","-","-",18,"*" "Other Renewable1",637,1.7,"3,181",2.3 "Petroleum","2,189",6.0,641,0.5 "Total","36,636",100.0,"137,577",100.0

236

Tennessee Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,401",15.9,"27,739",33.7 "Coal","8,805",41.1,"43,670",53.0 "Hydro and Pumped Storage","4,277",20.0,"7,416",9.0 "Natural Gas","4,655",21.7,"2,302",2.8 "Other1","-","-",16,"*" "Other Renewable1",222,1.0,988,1.2 "Petroleum",58,0.3,217,0.3 "Total","21,417",100.0,"82,349",100.0

237

Study of feasible and sustainable multilateral approach on nuclear fuel cycle  

SciTech Connect (OSTI)

Despite the Fukushima accident it is undeniable that nuclear power remains one of the most important methods to handle global growth of economic/energy consumption and issues with greenhouse gases. If the demand for nuclear power increases, the demand for not only the generation of power but also for refining uranium (U), conversion, enrichment, re-conversion, and fuel manufacturing should increase. In addition, concerns for the proliferation of 'Sensitive Nuclear Technologies' (SNT) should also increase. We propose a demand-side approach, where nuclear fuel cycle (NFC) activities would be implemented among multiple states. With this approach, NFC services, in particular those using SNTs, are multilaterally executed and controlled, thereby preventing unnecessary proliferation of SNTs, and enabling safe and appropriate control of nuclear technologies and nuclear materials. This proposal would implement nuclear safety and security at an international level and solve transport issues for nuclear fuels. This proposal is based on 3 types of cooperation for each element of NFC: type A: cooperation for 3S only, services received; Type B: cooperation for 3S, MNA (Multilateral Nuclear Activities) without transfer of ownership to MNA; and Type C cooperation for 3S, MNA holding ownership rights. States involved in the 3 types of activity should be referred to as partner states, host states, and site states respectively. The feasibility of the proposal is discussed for the Asian region.

Kuno, Y.; Tazaki, M. [University of Tokyo, Tokyo (Japan); Japan Atomic Energy Agency - JAEA, 4-49 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1184 (Japan); Akiba, M.; Takashima, R.; Izumi, Y.; Tanaka, S. [University of Tokyo, Tokyo (Japan)

2013-07-01T23:59:59.000Z

238

Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites |  

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

Evaluation of Removing Used Nuclear Fuel From Shutdown Evaluation of Removing Used Nuclear Fuel From Shutdown Sites Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites In January 2013, the Department of Energy issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America's Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses. Shutdown sites are defined as those commercial nuclear power reactor sites where the

239

SORPTION OF URANIUM, PLUTONIUM AND NEPTUNIUM ONTO SOLIDS PRESENT IN HIGH CAUSTIC NUCLEAR WASTE STORAGE TANKS  

SciTech Connect (OSTI)

Solids such as granular activated carbon, hematite and sodium phosphates, if present as sludge components in nuclear waste storage tanks, have been found to be capable of precipitating/sorbing actinides like plutonium, neptunium and uranium from nuclear waste storage tank supernatant liqueur. Thus, the potential may exists for the accumulation of fissile materials in such nuclear waste storage tanks during lengthy nuclear waste storage and processing. To evaluate the nuclear criticality safety in a typical nuclear waste storage tank, a study was initiated to measure the affinity of granular activated carbon, hematite and anhydrous sodium phosphate to sorb plutonium, neptunium and uranium from alkaline salt solutions. Tests with simulated and actual nuclear waste solutions established the affinity of the solids for plutonium, neptunium and uranium upon contact of the solutions with each of the solids. The removal of plutonium and neptunium from the synthetic salt solution by nuclear waste storage tank solids may be due largely to the presence of the granular activated carbon and transition metal oxides in these storage tank solids or sludge. Granular activated carbon and hematite also showed measurable affinity for both plutonium and neptunium. Sodium phosphate, used here as a reference sorbent for uranium, as expected, exhibited high affinity for uranium and neptunium, but did not show any measurable affinity for plutonium.

Oji, L; Bill Wilmarth, B; David Hobbs, D

2008-05-30T23:59:59.000Z

240

Uranium industry annual 1995  

SciTech Connect (OSTI)

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

NONE

1996-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

A Critical Step Toward Sustainable Nuclear Fuel Disposal | Department of  

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

A Critical Step Toward Sustainable Nuclear Fuel Disposal A Critical Step Toward Sustainable Nuclear Fuel Disposal A Critical Step Toward Sustainable Nuclear Fuel Disposal January 26, 2012 - 2:30pm Addthis Secretary Chu Secretary Chu Former Secretary of Energy The Blue Ribbon Commission on America's Nuclear Future was formed at the direction of the President to conduct a comprehensive review of polices for managing the back end of the nuclear fuel cycle. If we are going to ensure that the United States remains at the forefront of nuclear safety and security, non-proliferation, and nuclear energy technology we must develop an effective strategy and workable plan for the safe and secure management and disposal of used nuclear fuel and nuclear waste. That is why I asked General Scowcroft and Representative Hamilton to draw on their

242

Mr. William f. Crow, Acting Director . Uranium Fuel Licensing...  

Office of Legacy Management (LM)

Director Division of Facility and Site Decomnlssioning Projects Office of Nuclear Energy 3 Enclosures , ---- - . . : . ' . . . . ENCLOSURE 7 FORMERLY UTILIZED SITES...

243

Nuclear Fuels Storage & Transportation Planning Project | Department of  

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

Nuclear Fuels Storage & Nuclear Fuels Storage & Transportation Planning Project Nuclear Fuels Storage & Transportation Planning Project Independent Spent Fuel Storage Installation (ISFSI) at the shutdown Connecticut Yankee site. The ISFSI includes 40 multi-purpose canisters, within vertical concrete storage casks, containing 1019 used nuclear fuel assemblies [412.3 metric ton heavy metal (MTHM)] and 3 canisters of greater-than-class-C (GTCC) low-level radioactive waste. Photo courtesy of Connecticut Yankee (http://www.connyankee.com/html/fuel_storage.html). Independent Spent Fuel Storage Installation (ISFSI) at the shutdown Connecticut Yankee site. The ISFSI includes 40 multi-purpose canisters, within vertical concrete storage casks, containing 1019 used nuclear fuel

244

Uranium: Environmental Pollution and Health Effects  

Science Journals Connector (OSTI)

Uranium is found ubiquitously in nature in low concentrations in soil, rock, and water. Naturally occurring uranium contains three isotopes, namely 238U, 235U, and 234U. All uranium isotopes have the same chemical properties, but they have different radiological properties. The main civilian use of uranium is to fuel nuclear power plants, whereas high enriched (in 235U) uranium is used in the military sector as nuclear explosives and depleted uranium (DU) as penetrators or tank shielding. Exposure to uranium may cause health problems due to its radiological (uranium is predominantly emitting alpha-particles) and chemical actions (heavy metal toxicity). Uranium uptake may occur by ingestion, inhalation, contaminated wounds, and embedded fragments especially for soldiers. Inhalation of dust is considered the major pathway for uranium uptake in workplaces. Soluble uranium compounds tend to quickly pass through the body, whereas insoluble uranium compounds pose a more serious inhalation exposure hazard. The kidney is the most sensitive organ for uranium chemotoxicity. An important indirect radiological effect of uranium is the increased risk of lung cancers from inhalation of the daughter products of radon, a noble gas in the uranium decay chains that transports uranium-derived radioactivity from soil into the indoor environment. No direct evidence about the carcinogenic effect of DU in humans is available yet.

D. Melo; W. Burkart

2011-01-01T23:59:59.000Z

245

EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering  

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

EIS-0203: Spent Nuclear Fuel Management and Idaho National EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs SUMMARY This EIS considers programmatic (DOE-wide) alternative approaches to safely, efficiently, and responsibly manage existing and projected quantities of spent nuclear fuel until the year 2035. This amount of time may be required to make and implement a decision on the ultimate disposition of spent nuclear fuel. DOE's spent nuclear fuel responsibilities include fuel generated by DOE production, research, and development reactors; naval reactors; university and foreign research reactors; domestic non-DOE reactors such as those at the National Institute

246

Transportation capabilities study of DOE-owned spent nuclear fuel  

SciTech Connect (OSTI)

This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1994-10-01T23:59:59.000Z

247

The Rhode Island Nuclear Science Center conversion from HEU to LEU fuel  

SciTech Connect (OSTI)

The 2-MW Rhode Island Nuclear Science Center (RINSC) open pool reactor was converted from 93% UAL-High Enriched Uranium (HEU) fuel to 20% enrichment U3Si2-AL Low Enriched Uranium (LEU) fuel. The conversion included redesign of the core to a more compact size and the addition of beryllium reflectors and a beryllium flux trap. A significant increase in thermal flux level was achieved due to greater neutron leakage in the new compact core configuration. Following the conversion, a second cooling loop and an emergency core cooling system were installed to permit operation at 5 MW. After re-licensing at 2 MW, a power upgrade request will be submitted to the NRC.

Tehan, Terry

2000-09-27T23:59:59.000Z

248

Dynamic Systems Analysis Report for Nuclear Fuel Recycle  

SciTech Connect (OSTI)

This report examines the time-dependent dynamics of transitioning from the current United States (U.S.) nuclear fuel cycle where used nuclear fuel is disposed in a repository to a closed fuel cycle where the used fuel is recycled and only fission products and waste are disposed. The report is intended to help inform policy developers, decision makers, and program managers of system-level options and constraints as they guide the formulation and implementation of advanced fuel cycle development and demonstration efforts and move toward deployment of nuclear fuel recycling infrastructure.

Brent Dixon; Sonny Kim; David Shropshire; Steven Piet; Gretchen Matthern; Bill Halsey

2008-12-01T23:59:59.000Z

249

Assessment of National Nuclear Fuel Cycle for Transmutations of High Level Nuclear Waste  

Science Journals Connector (OSTI)

The advanced fuel cycle initiative (AFCI) has been investigated for the safe processing of the spent nuclear fuels (SNFs), which has focused mainly ... of the SNFs considering the characteristics of the nuclear m...

Taeho Woo

2012-01-01T23:59:59.000Z

250

Uranium Management and Policy | Department of Energy  

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

Test Program, and reporting annually to Congress on the impact of the U.S.-Russia Highly Enriched Uranium Purchase Agreement on the U.S. nuclear fuel industry. NE-54's...

251

Proceedings of GLOBAL 2013: International Nuclear Fuel Cycle Conference - Nuclear Energy at a Crossroads  

SciTech Connect (OSTI)

The Global conference is a forum for the discussion of the scientific, technical, social and regulatory aspects of the nuclear fuel cycle. Relevant topics include global utilization of nuclear energy, current fuel cycle technologies, advanced reactors, advanced fuel cycles, nuclear nonproliferation and public acceptance.

NONE

2013-07-01T23:59:59.000Z

252

Examination of the conversion of the U.S. submarine fleet from highly enriched uranium to low enriched uranium ; Examination of the conversion of the United States submarine fleet from HEU to low LEU .  

E-Print Network [OSTI]

??The nuclear reactors used by the U.S. Navy for submarine propulsion are currently fueled by highly enriched uranium (HEU), but HEU brings administrative and political… (more)

McCord, Cameron (Cameron Liam)

2014-01-01T23:59:59.000Z

253

Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14  

SciTech Connect (OSTI)

The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

Schneider, K.J.

1982-09-01T23:59:59.000Z

254

Pilot Application to Nuclear Fuel Cycle Options | Department of Energy  

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

Pilot Application to Nuclear Fuel Cycle Options Pilot Application to Nuclear Fuel Cycle Options Pilot Application to Nuclear Fuel Cycle Options A Screening Method for Guiding R&D Decisions: Pilot Application to Screen Nuclear Fuel Cycle Options The Department of Energy's Office of Nuclear Energy (DOE-NE) invests in research and development (R&D) to ensure that the United States will maintain its domestic nuclear energy capability and scientific and technical leadership in the international community of nuclear power nations in the years ahead. The 2010 Nuclear Energy Research and Development Roadmap presents a high-level vision and framework for R&D activities that are needed to keep the nuclear energy option viable in the near term and to expand its use in the decades ahead. The roadmap identifies the development

255

Fuel assembly transfer basket for pool type nuclear reactor vessels  

DOE Patents [OSTI]

A fuel assembly transfer basket for a pool type, liquid metal cooled nuclear reactor having a side access loading and unloading port for receiving and relinquishing fuel assemblies during transfer.

Fanning, Alan W. (San Jose, CA); Ramsour, Nicholas L. (San Jose, CA)

1991-01-01T23:59:59.000Z

256

Microsoft Word - spent nuclear fuel report.doc  

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

Management of Spent Nuclear Fuel Management of Spent Nuclear Fuel at the Savannah River Site DOE/IG-0727 May 2006 REPORT ON MANAGEMENT OF SPENT NUCLEAR FUEL AT THE SAVANNAH RIVER SITE TABLE OF CONTENTS Spent Nuclear Fuel Management Details of Finding 1 Recommendations 2 Comments 3 Appendices 1. Objective, Scope, and Methodology 4 2. Prior Audit Reports 5 3. Management Comments 6 SPENT NUCLEAR FUEL MANGEMENT Page 1 Details of Finding H-Canyon The Department of Energy's (Department) spent nuclear fuel Operations program at the Savannah River Site (Site) will likely require Extended H-Canyon to be maintained at least two years beyond defined operational needs. The Department committed to maintain H-Canyon operational readiness to provide a disposal path for

257

EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina | Department  

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

79: Spent Nuclear Fuel Management, Aiken, South Carolina 79: Spent Nuclear Fuel Management, Aiken, South Carolina EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina SUMMARY The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD April 5, 2013 EIS-0279: Amended Record of Decision Spent Nuclear Fuel Management at the Savannah River Site April 1, 2013 EIS-0279-SA-01: Supplement Analysis Savannah River Site Spent Nuclear Fuel Management (DOE/EIS-0279-SA-01 and

258

Nuclear reactor fuel rod attachment system  

DOE Patents [OSTI]

A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

Christiansen, David W. (Kennewick, WA)

1982-01-01T23:59:59.000Z

259

LANL disassembles "pits," makes mixed-oxide fuel  

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

at the MOX facility in South Carolina, the plutonium oxide from LANL will be blended with depleted uranium, fabricated into MOX fuel, and irradiated in domestic nuclear...

260

Energy Department Announces New Investment in Nuclear Fuel Storage Research  

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

Investment in Nuclear Fuel Storage Investment in Nuclear Fuel Storage Research Energy Department Announces New Investment in Nuclear Fuel Storage Research April 16, 2013 - 12:19pm Addthis NEWS MEDIA CONTACT (202) 586-4940 WASHINGTON - As part of its commitment to developing an effective strategy for the safe and secure storage and management of used nuclear fuel, the Energy Department today announced a new dry storage research and development project led by the Electric Power Research Institute (EPRI). The project will design and demonstrate dry storage cask technology for high burn-up spent nuclear fuels that have been removed from commercial nuclear power plants. "The Energy Department is committed to advancing clean, reliable and safe nuclear power - which provides the largest source of low-carbon

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Energy Department Announces New Investment in Nuclear Fuel Storage Research  

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

Announces New Investment in Nuclear Fuel Storage Announces New Investment in Nuclear Fuel Storage Research Energy Department Announces New Investment in Nuclear Fuel Storage Research April 16, 2013 - 12:19pm Addthis NEWS MEDIA CONTACT (202) 586-4940 WASHINGTON - As part of its commitment to developing an effective strategy for the safe and secure storage and management of used nuclear fuel, the Energy Department today announced a new dry storage research and development project led by the Electric Power Research Institute (EPRI). The project will design and demonstrate dry storage cask technology for high burn-up spent nuclear fuels that have been removed from commercial nuclear power plants. "The Energy Department is committed to advancing clean, reliable and safe nuclear power - which provides the largest source of low-carbon

262

DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER  

SciTech Connect (OSTI)

The DOE Spent Nuclear Fuel Disposal Container (SNF DC) supports the confinement and isolation of waste within the Engineered Barrier System of the Mined Geologic Disposal System (MGDS). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the access mains, and emplaced in emplacement drifts. The DOE Spent Nuclear Fuel Disposal Container provides long term confinement of DOE SNF waste, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The DOE SNF Disposal Containers provide containment of waste for a designated period of time, and limit radionuclide release thereafter. The disposal containers maintain the waste in a designated configuration, withstand maximum handling and rockfall loads, limit the individual waste canister temperatures after emplacement. The disposal containers also limit the introduction of moderator into the disposal container during the criticality control period, resist corrosion in the expected repository environment, and provide complete or limited containment of waste in the event of an accident. Multiple disposal container designs may be needed to accommodate the expected range of DOE Spent Nuclear Fuel. The disposal container will include outer and inner barrier walls and outer and inner barrier lids. Exterior labels will identify the disposal container and contents. Differing metal barriers will support the design philosophy of defense in depth. The use of materials with different failure mechanisms prevents a single mode failure from breaching the waste package. The corrosion-resistant inner barrier and inner barrier lid will be constructed of a high-nickel alloy and the corrosion-allowance outer barrier and outer barrier lid will be made of carbon steel. The DOE Spent Nuclear Fuel Disposal Containers interface with the emplacement drift environment by transferring heat from the waste to the external environment and by protecting the DOE waste canisters and their contents from damage/degradation by the external environment. The disposal containers also interface with the SNF by limiting access of moderator and oxidizing agents to the waste. The disposal containers interface with the Ex-Container System's emplacement drift disposal container supports. The disposal containers interface with the Canister Transfer System, Waste Emplacement System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and remediation of the disposal container.

F. Habashi

1998-06-26T23:59:59.000Z

263

Used Nuclear Fuel Loading and Structural Performance Under Normal...  

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

deterministic evaluations of moderate-to-high burnup used nuclear fuel (UNF) mechanical performance under normal conditions of storage (NCS) and normal conditions of...

264

Spent nuclear fuel project - criteria document spent nuclear fuel final safety analysis report  

SciTech Connect (OSTI)

The criteria document provides the criteria and planning guidance for developing the Spent Nuclear Fuel (SNF) Final Safety Analysis Report (FSAR). This FSAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, installation acceptance testing, startup, and operation of the SNF Project facilities (K Basins, Cold Vacuum Drying Facility, and Canister Storage Building).

MORGAN, R.G.

1999-02-23T23:59:59.000Z

265

Uranium and nuclear power: The role of exploration information in framing public policy  

Science Journals Connector (OSTI)

Abstract As the importance of addressing climate change increases the future global role for nuclear power, the demand for uranium will increase. Expanded uranium reserves will be needed to meet this increased demand, highlighting the importance of future exploratory efforts. To shed light on the social desirability of future exploration levels I analyze a past expansionary period in the U.S. uranium industry. I find exploration levels were smaller than socially efficient during this period, resulting from a deviation between the private and social values of information. Looking forward, public policies can encourage optimal exploration levels by addressing this deviation.

Charles F. Mason

2014-01-01T23:59:59.000Z

266

Environmental assessment for the purchase of Russian low enriched uranium derived from the dismantlement of nuclear weapons in the countries of the former Soviet Union  

SciTech Connect (OSTI)

The United States is proposing to purchase from the Russian Federation low enriched uranium (LEU) derived from highly enriched uranium (HEU) resulting from the dismantlement of nuclear weapons in the countries of the former Soviet Union. The purchase would be accomplished through a proposed contract requiring the United States to purchase 15,250 metric tons (tonnes) of LEU (or 22,550 tonnes of UF{sub 6}) derived from blending 500 metric tones uranium (MTU) of HEU from nuclear warheads. The LEU would be in the form of uranium hexafluoride (UF{sub 6}) and would be converted from HEU in Russia. The United States Enrichment Corporation (USEC) is the entity proposing to undertake the contract for purchase, sale, and delivery of the LEU from the Russian Federation. The US Department of Energy (DOE) is negotiating the procedure for gaining confidence that the LEU is derived from HEU that is derived from dismantled nuclear weapons (referred to as ``transparency),`` and would administer the transparency measures for the contract. There are six environments that could potentially be affected by the proposed action; marine (ocean); US ports of entry; truck or rail transportation corridors; the Portsmouth GDP; the electric power industry; and the nuclear fuel cycle industry. These environmental impacts are discussed.

Not Available

1994-01-01T23:59:59.000Z

267

Risk and Responsibility Sharing in Nuclear Spent Fuel Management  

E-Print Network [OSTI]

With the Nuclear Waste Policy Act of 1982, the responsibility of American utilities in the long-term management of spent nuclear fuel was limited to the payment of a fee. This narrow involvement did not result in faster ...

De Roo, Guillaume

268

FAQ 3-What are the common forms of uranium?  

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

are the common forms of uranium? are the common forms of uranium? What are the common forms of uranium? Uranium can take many chemical forms. In nature, uranium is generally found as an oxide, such as in the olive-green-colored mineral pitchblende. Uranium oxide is also the chemical form most often used for nuclear fuel. Uranium-fluorine compounds are also common in uranium processing, with uranium hexafluoride (UF6) and uranium tetrafluoride (UF4) being the two most common. In its pure form, uranium is a silver-colored metal. The most common forms of uranium oxide are U3O8 and UO2. Both oxide forms have low solubility in water and are relatively stable over a wide range of environmental conditions. Triuranium octaoxide (U3O8) is the most stable form of uranium and is the form most commonly found in nature. Uranium dioxide (UO2) is the form in which uranium is most commonly used as a nuclear reactor fuel. At ambient temperatures, UO2 will gradually convert to U3O8. Because of their stability, uranium oxides are generally considered the preferred chemical form for storage or disposal.

269

Review of Used Nuclear Fuel Storage and Transportation Technical Gap  

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

Analyses Analyses Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analyses The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation, and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The Storage and Transportation activities within the UFDC are being developed to address issues regarding the extended storage of UNF and its subsequent

270

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

271

Modeling of Gap Closure in Uranium-Zirconium Alloy Metal Fuel - A Test Problem  

SciTech Connect (OSTI)

Uranium based binary and ternary alloy fuel is a possible candidate for advanced fast spectrum reactors with long refueling intervals and reduced liner heat rating [1]. An important metal fuel issue that can impact the fuel performance is the fuel-cladding gap closure, and fuel axial growth. The dimensional change in the fuel during irradiation is due to a superposition of the thermal expansion of the fuel due to heating, volumetric changes due to possible phase transformations that occur during heating and the swelling due to fission gas retention. The volumetric changes due to phase transformation depend both on the thermodynamics of the alloy system and the kinetics of phase change reactions that occur at the operating temperature. The nucleation and growth of fission gas bubbles that contributes to fuel swelling is also influenced by the local fuel chemistry and the microstructure. Once the fuel expands and contacts the clad, expansion in the radial direction is constrained by the clad, and the overall deformation of the fuel clad assembly depends upon the dynamics of the contact problem. The neutronics portion of the problem is also inherently coupled with microstructural evolution in terms of constituent redistribution and phase transformation. Because of the complex nature of the problem, a series of test problems have been defined with increasing complexity with the objective of capturing the fuel-clad interaction in complex fuels subjected to a wide range of irradiation and temperature conditions. The abstract, if short, is inserted here before the introduction section. If the abstract is long, it should be inserted with the front material and page numbered as such, then this page would begin with the introduction section.

Simunovic, Srdjan [ORNL; Ott, Larry J [ORNL; Gorti, Sarma B [ORNL; Nukala, Phani K [ORNL; Radhakrishnan, Balasubramaniam [ORNL; Turner, John A [ORNL

2009-10-01T23:59:59.000Z

272

Methodology for comparing the health effects of electricity generation from uranium and coal fuels  

SciTech Connect (OSTI)

A methodology was developed for comparing the health risks of electricity generation from uranium and coal fuels. The health effects attributable to the construction, operation, and decommissioning of each facility in the two fuel cycle were considered. The methodology is based on defining (1) requirement variables for the materials, energy, etc., (2) effluent variables associated with the requirement variables as well as with the fuel cycle facility operation, and (3) health impact variables for effluents and accidents. The materials, energy, etc., required for construction, operation, and decommissioning of each fuel cycle facility are defined as primary variables. The materials, energy, etc., needed to produce the primary variable are defined as secondary requirement variables. Each requirement variable (primary, secondary, etc.) has associated effluent variables and health impact variables. A diverging chain or tree is formed for each primary variable. Fortunately, most elements reoccur frequently to reduce the level of analysis complexity. 6 references, 11 figures, 6 tables.

Rhyne, W.R.; El-Bassioni, A.A.

1981-12-08T23:59:59.000Z

273

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel  

SciTech Connect (OSTI)

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.

Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL

2014-10-30T23:59:59.000Z

274

Idaho site completes demolition of Cold War-era nuclear fuel reprocessing  

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

doe logo CH2M-WG logo Joint News Release For Immediate Release Date: December 22, 2011 Media Contact: Erik Simpson, (208) 360-0426 Idaho site completes demolition of Cold War-era nuclear fuel reprocessing facility A gravel mound, larger than half a city block and several feet thick, is the only visible feature that remains at the site of a Cold War-era spent nuclear fuel reprocessing facility at the U.S. Department of Energy�s Idaho site. About $44 million in American Recovery and Reinvestment Act funds helped Idaho Cleanup Project crews accelerate the demolition of the facility that during its 40 years of operation recovered more than $1 billion worth of uranium. �The ability to retain our highly skilled workforce was a huge contributor to the success of this project,� said Idaho Cleanup Project

275

Experimental Investigation of Burnup Credit for Safe Transport, Storage, and Disposal of Spent Nuclear Fuel  

SciTech Connect (OSTI)

This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

Harms, Gary A.; Helmick, Paul H.; Ford, John T.; Walker, Sharon A.; Berry, Donald T.; Pickard, Paul S.

2004-04-01T23:59:59.000Z

276

The Chemistry os Spent Nuclear Fuel From X-Ray Absorption Spectroscopy  

SciTech Connect (OSTI)

Present and future nuclear fuel cycles will require an understanding of the complex chemistry of trace fission products and transuranium actinides in spent nuclear fuel (SNF). Because of the unique analytical challenges presented by SNF to the materials scientist, many of its fundamental physical and chemical properties remain poorly understood, especially on the microscopic scale. Such an understanding of the chemical states of radionuclides in SNF would benefit development of technologies for fuel monitoring, fuel performance improvement and modeling, fuel reprocessing, and spent fuel storage and disposal. We have recently demonstrated the use of synchrotron x-ray absorption spectroscopy (XAS) to examine crystal chemical properties of actinides and fission products in extracted specimens of SNF. Information obtained includes oxidation state, chemical bond coordination, and quantitative elemental concentration and distribution. We have also used XAS in a scanning mode to obtain x-ray spectral micrographs with resolution approaching 1 micron. A brief overview of the technique will be presented, along with findings on uranium, plutonium, neptunium, technetium, and molybdenum in commercial PWR SNF specimens.

F.A. Fortner; A.J. Kropf; J.C. Cunnane

2006-09-21T23:59:59.000Z

277

A Novel Fuel/Reactor Cycle to Implement the 300 Years Nuclear Waste Policy Approach - 12377  

SciTech Connect (OSTI)

A thorium-based fuel cycle system can effectively burn the currently accumulated commercial used nuclear fuel and move to a sustainable equilibrium where the actinide levels in the high level waste are low enough to yield a radiotoxicity after 300 years lower than that of the equivalent uranium ore. The second step of the Westinghouse approach to solving the waste 'problem' has been completed. The thorium fuel cycle has indeed the potential of burning the legacy TRU and achieve the waste objective proposed. Initial evaluations have been started for the third step, development and selection of appropriate reactors. Indications are that the probability of show-stoppers is rather remote. It is, therefore, believed that development of the thorium cycle and associated technologies will provide a permanent solution to the waste management. Westinghouse is open to the widest collaboration to make this a reality. (authors)

Carelli, M.D.; Franceschini, F.; Lahoda, E.J. [Westinghouse Electric Company LLC., Cranberry Township, PA (United States); Petrovic, B. [Georgia Institute of Technology, Atlanta, GA (United States)

2012-07-01T23:59:59.000Z

278

Used Nuclear Fuel Loading and Structural Performance Under Normal  

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

Nuclear Fuel Loading and Structural Performance Under Normal Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Demonstration of Approach and Results of Used Fuel Performance Characterization Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Demonstration of Approach and Results of Used Fuel Performance Characterization This report provides results of the initial demonstration of the modeling capability developed to perform preliminary deterministic evaluations of moderate-to-high burnup used nuclear fuel (UNF) mechanical performance under normal conditions of storage (NCS) and normal conditions of transport (NCT) conditions. This report also provides results from the sensitivity studies, and discussion on the long-term goals and objectives of this

279

Technology Readiness Levels for Advanced Nuclear Fuels and Materials Development  

SciTech Connect (OSTI)

The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Advanced nuclear fuels and materials development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Advanced Fuel Campaign (AFC).

Jon Carmack

2014-01-01T23:59:59.000Z

280

Coupon Surveillance For Corrosion Monitoring In Nuclear Fuel Basin  

SciTech Connect (OSTI)

Aluminum and stainless steel coupons were put into a nuclear fuel basin to monitor the effect of water chemistry on the corrosion of fuel cladding. These coupons have been monitored for over ten years. The corrosion and pitting data is being used to model the kinetics and estimate the damage that is occurring to the fuel cladding.

Mickalonis, J. I.; Murphy, T. R.; Deible, R.

2012-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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
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281

Dry halide method for separating the components of spent nuclear fuels  

DOE Patents [OSTI]

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

Christian, Jerry Dale (Idaho Falls, ID); Thomas, Thomas Russell (Rigby, ID); Kessinger, Glen F. (Idaho Falls, ID)

1998-01-01T23:59:59.000Z

282

Standard test methods for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of uranium hexafluoride  

E-Print Network [OSTI]

1.1 These test methods cover procedures for subsampling and for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of uranium hexafluoride UF6. Most of these test methods are in routine use to determine conformance to UF6 specifications in the Enrichment and Conversion Facilities. 1.2 The analytical procedures in this document appear in the following order: Note 1—Subcommittee C26.05 will confer with C26.02 concerning the renumbered section in Test Methods C761 to determine how concerns with renumbering these sections, as analytical methods are replaced with stand-alone analytical methods, are best addressed in subsequent publications. Sections Subsampling of Uranium Hexafluoride 7 - 10 Gravimetric Determination of Uranium 11 - 19 Titrimetric Determination of Uranium 20 Preparation of High-Purity U3O 8 21 Isotopic Analysis 22 Isotopic Analysis by Double-Standard Mass-Spectrometer Method 23 - 29 Determination of Hydrocarbons, Chlorocarbons, and Partially Substitut...

American Society for Testing and Materials. Philadelphia

2011-01-01T23:59:59.000Z

283

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel  

SciTech Connect (OSTI)

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

Cowell, B.S.; Fisher, S.E.

1999-02-01T23:59:59.000Z

284

New Hampshire Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,247",29.8,"10,910",49.2 "Coal",546,13.1,"3,083",13.9 "Hydro and Pumped Storage",489,11.7,"1,478",6.7 "Natural Gas","1,215",29.1,"5,365",24.2 "Other1","-","-",57,0.3 "Other Renewable1",182,4.4,"1,232",5.6 "Petroleum",501,12.0,72,0.3 "Total","4,180",100.0,"22,196",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

285

New Jersey Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,108",22.3,"32,771",49.9 "Coal","2,036",11.1,"6,418",9.8 "Hydro and Pumped Storage",404,2.2,-176,-0.3 "Natural Gas","10,244",55.6,"24,902",37.9 "Other1",56,0.3,682,1.0 "Other Renewable1",226,1.2,850,1.3 "Petroleum","1,351",7.3,235,0.4 "Total","18,424",100.0,"65,682",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

286

January 2013 Most Viewed Documents for Fission And Nuclear Technologie...  

Office of Scientific and Technical Information (OSTI)

January 2013 Most Viewed Documents for Fission And Nuclear Technologies Laboratory studies of shearleach processing of zircaloy clad metallic uranium reactor fuel Swanson, J.L.;...

287

Benefits and concerns of a closed nuclear fuel cycle  

SciTech Connect (OSTI)

Nuclear power can play an important role in our energy future, contributing to increasing electricity demand while at the same time decreasing carbon dioxide emissions. However, the nuclear fuel cycle in the United States today is unsustainable. As stated in the 1982 Nuclear Waste Policy Act, the U.S. Department of Energy is responsible for disposing of spent nuclear fuel generated by commercial nuclear power plants operating in a “once-through” fuel cycle in the deep geologic repository located at Yucca Mountain. However, unyielding political opposition to the site has hindered the commissioning process to the extant that the current administration has recently declared the unsuitability of the Yucca Mountain site. In light of this the DOE is exploring other options, including closing the fuel cycle through recycling and reprocessing of spent nuclear fuel. The possibility of closing the fuel cycle is receiving special attention because of its ability to minimize the final high level waste (HLW) package as well as recover additional energy value from the original fuel. The technology is, however, still very controversial because of the increased cost and proliferation risk it can present. To lend perspective on the closed fuel cycle alternative, this presents the arguments for and against closing the fuel cycle with respect to sustainability, proliferation risk, commercial viability, waste management, and energy security.

Widder, Sarah H.

2010-11-17T23:59:59.000Z

288

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network [OSTI]

Security of the National Nuclear Security Administration, USof Energys National Nuclear Security Administration (NNSA)

Quiter, Brian

2012-01-01T23:59:59.000Z

289

Y-12 Knows Uranium | Y-12 National Security Complex  

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

Knows Uranium Knows Uranium Y-12 Knows Uranium Posted: July 22, 2013 - 3:45pm | Y-12 Report | Volume 10, Issue 1 | 2013 Y-12 produces many forms of uranium. They may be used in chemical processing steps on-site or shipped elsewhere to serve as raw materials for nuclear fuel or as research tools. All of uranium's uses, defense related and otherwise, are critical to the nation. Y-12's understanding of uranium, coupled with the site's work with enriched uranium metal, alloys, oxides, compounds and solutions, is unique in the Nuclear Security Enterprise. "The Y-12 work force understands both established uranium science and the esoteric things related to uranium's behavior," said engineer Alan Moore. "Such a deep, detailed understanding comes from experience,

290

Assessment of uranium-free nitride fuels for spent fuel transmutation in fast reactor systems  

E-Print Network [OSTI]

. ....................................................................................... 18 Fig. 4. Standard PWR ¼ core model with fresh, once- and twice-burned fuel, and the location of MOX fuel assemblies with respect to original layout, 32% MOX loading................................................................................................................ 21 Fig. 5. Control rod locations......................................................................................... 21 Fig. 6. Net change of U, Pu and Am for PWR and 1/3 MOX fueled whole cores, 360 day burn...

Szakaly, Frank Joseph

2004-09-30T23:59:59.000Z

291

A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods  

Science Journals Connector (OSTI)

Abstract Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (?,n) interrogation source strength of 5.7×104 s?1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32 gadolinia burnable poison rods with Gd concentrations of up to 12 wt%. Monte Carlo calculations predict that the EFC has a lower statistical uncertainty for measurements performed in the fast neutron mode than its predecessor neutron collar design. This paper describes the physics design and calculated performance characteristics of the EFC. The Gd response is presented over a realistic range for modern PWR fuel designs.

Louise G. Evans; Martyn T. Swinhoe; Howard O. Menlove; Peter Schwalbach; Paul De Baere; Michael C. Browne

2013-01-01T23:59:59.000Z

292

Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2011-05-01T23:59:59.000Z

293

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

SciTech Connect (OSTI)

In nuclear resonance fluorescence (NRF) measurements, resonances are excited by an external photon beam leading to the emission of gamma rays with specific energies that are characteristic of the emitting isotope. NRF promises the unique capability of directly quantifying a specific isotope without the need for unfolding the combined responses of several fissile isotopes as is required in other measurement techniques. We have analyzed the potential of NRF as a non-destructive analysis technique for quantitative measurements of Pu isotopes in spent nuclear fuel (SNF). Given the low concentrations of 239Pu in SNF and its small integrated NRF cross sections, the main challenge in achieving precise and accurate measurements lies in accruing sufficient counting statistics in a reasonable measurement time. Using analytical modeling, and simulations with the radiation transport code MCNPX that has been experimentally tested recently, the backscatter and transmission methods were quantitatively studied for differing photon sources and radiation detector types. Resonant photon count rates and measurement times were estimated for a range of photon source and detection parameters, which were used to determine photon source and gamma-ray detector requirements. The results indicate that systems based on a bremsstrahlung source and present detector technology are not practical for high-precision measurements of 239Pu in SNF. Measurements that achieve the desired uncertainties within hour-long measurements will either require stronger resonances, which may be expressed by other Pu isotopes, or require quasi-monoenergetic photon sources with intensities that are approximately two orders of magnitude higher than those currently being designed or proposed.This work is part of a larger effort sponsored by the Next Generation Safeguards Initiative to develop an integrated instrument, comprised of individual NDA techniques with complementary features, that is fully capable of determining Pu mass in spent fuel assemblies.

Quiter, Brian; Ludewigt, Bernhard; Ambers, Scott

2011-06-30T23:59:59.000Z

294

What to Expect When Readying to Move Spent Nuclear Fuel from...  

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

What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power...

295

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

296

Microstructural Examination to Aid in Understanding Friction Bonding Fabrication Technique for Monolithic Nuclear Fuel  

SciTech Connect (OSTI)

Monolithic nuclear fuel is currently being developed for use in research reactors, and friction bonding (FB) is a technique being developed to help in this fuel’s fabrication. Since both FB and monolithic fuel are new concepts, research is needed to understand the impact of varying FB fabrication parameters on fuel plate characteristics. This thesis research provides insight into the FB process and its application to the monolithic fuel design by recognizing and understanding the microstructural effects of varying fabrication parameters (a) FB tool load, and (b) FB tool face alloy. These two fabrication parameters help drive material temperature during fabrication, and thus the material properties, bond strength, and possible formation of interface reaction layers. This study analyzed temperatures and tool loads measured during those FB processes and examined microstructural characteristics of materials and bonds in samples taken from the resulting fuel plates. This study shows that higher tool load increases aluminum plasticization and forging during FB, and that the tool face alloy helps determine the tool’s heat extraction efficacy. The study concludes that successful aluminum bonds can be attained in fuel plates using a wide range of FB tool loads. The range of tool loads yielding successful uranium-aluminum bonding was not established, but it was demonstrated that such bonding can be attained with FB tool load of 48,900 N (11,000 lbf) when using a FB tool faced with a tungsten alloy. This tool successfully performed FB, and with better results than tools faced with other materials. Results of this study correlate well with results reported for similar aluminum bonding techniques. This study’s results also provide support and validation for other nuclear fuel development studies and conclusions. Recommendations are offered for further research.

Karen L. Shropshire

2008-04-01T23:59:59.000Z

297

Materials Modeling and Simulation for Nuclear Fuels (MMSNF) Workshops  

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

Aerial photo of Argonne National Laboratory Argonne National Laboratory University of Chicago Chicago Photography courtesy Thomas F Ewing Privacy and Security Notice The MMSNF Workshops The goal of the Materials Modeling and Simulation for Nuclear Fuels (MMSNF) workshops is to stimulate research and discussions on modeling and simulations of nuclear fuels, to assist the design of improved fuels and the evaluation of fuel performance. In addition to research focused on existing or improved types of LWR reactors, recent modeling programs, networks, and links have been created to develop innovative nuclear fuels and materials for future generations of nuclear reactors. Examples can be found in Europe (e.g. F-BRIDGE project and ACTINET network and SAMANTHA cooperative network), in the USA (e.g. CASL, NEAMS, CESAR and CMSN network

298

Used Nuclear Fuel Loading and Structural Performance Under Normal  

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

Nuclear Fuel Loading and Structural Performance Under Normal Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan Used nuclear fuel (UNF) must maintain its integrity during the storage period in such a way that it can withstand the physical forces of handling and transportation associated with restaging the fuel and transporting it to treatment or recycling facilities, or to a geologic repository. This RD&D plan describes a methodology, including development and use of analytical models, to evaluate loading and associated mechanical responses of UNF rods and key structural components. The plan objective is to

299

Used Nuclear Fuel Loading and Structural Performance Under Normal  

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

Used Nuclear Fuel Loading and Structural Performance Under Normal Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan Used nuclear fuel (UNF) must maintain its integrity during the storage period in such a way that it can withstand the physical forces of handling and transportation associated with restaging the fuel and transporting it to treatment or recycling facilities, or to a geologic repository. This RD&D plan describes a methodology, including development and use of analytical models, to evaluate loading and associated mechanical responses of UNF rods and key structural components. The plan objective is to

300

Nuclear fuel corrosion over millennia interpreted using geologic data  

SciTech Connect (OSTI)

Corrosion of nuclear fuel over the 10,000 year regulatory period in a geologic repository will be a function of physical characteristics (e.g., crystallinity, crystal sizes, crystal forms) and chemical characteristics (e.g., crystal composition, compositional variability, accessory phases). Natural uraninite (nominally UO{sub 2+x}) which has undergone long-term corrosion can be studied to infer the long-term behavior of nuclear fuel. Previously, uraninite from the Nopal I deposit, Pena Blanca district, Chihuahua, Mexico, has been shown to constitute an outstanding analog material for comparison with nuclear fuel. Similarities between Nopal I uraninite and nuclear fuel have been shown to include bulk composition, general crystal structure, and total trace element content. Data presented here suggest that, as a bulk material, Nopal I uraninite compares favorably with irradiated nuclear fuel. Nevertheless, some fine-scale differences are noted between Nopal I uraninite and irradiated nuclear fuel with respect to both internal structures and compositions. These observations suggest that whereas the long-term responses of the two materials to oxidative alteration in a geologic repository may be similar, the detailed mechanisms of initial oxidant penetration and the short-term oxidative alternation of Nopal I uraninite and irradiated nuclear fuel are likely to be different.

Pearcy, E.C.; Manaktala, H.K. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

1994-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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 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

302

Laser-based characterization of nuclear fuel plates  

SciTech Connect (OSTI)

Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

Smith, James A.; Cottle, Dave L.; Rabin, Barry H. [Idaho National Laboratory, Fuel Performance and Design, P.O. Box 1625, Idaho Falls, Idaho, 83415-6188 (United States)

2014-02-18T23:59:59.000Z

303

Ukraine Loads U.S. Nuclear Fuel into Power Plant as Part of DOE-Ukraine Nuclear Fuel Qualification Program  

Broader source: Energy.gov [DOE]

fficials from the U.S. Department of Energy’s (DOE) Office of Nuclear Energy today (April 8, 2010) participated in a ceremony in Ukraine to mark the insertion of Westinghouse-produced nuclear fuel into a nuclear power plant in Ukraine.

304

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

305

Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering Division  

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

Analysis Analysis Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Overview Current Projects Software Nuclear Plant Dynamics and Safety Nuclear Data Program Advanced Reactor Development Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Reactor Physics and Fuel Cycle Analysis Bookmark and Share Reactor physics and fuel cycle analysis is a core competency of the Nuclear Engineering (NE) Division. The Division has played a major role in the design and analysis of advanced reactors, particularly liquid-metal-cooled reactors. NE researchers have concentrated on developing computer codes for

306

Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining  

SciTech Connect (OSTI)

A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl – 1 wt% Li2O at 650 °C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 °C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

S. D. Herrmann; S. X. Li

2010-09-01T23:59:59.000Z

307

Centrifugal microfluidic platform for radiochemistry: Potentialities for the chemical analysis of nuclear spent fuels  

Science Journals Connector (OSTI)

Abstract The use of a centrifugal microfluidic platform is for the first time reported as an alternative to classical chromatographic procedures for radiochemistry. The original design of the microfluidic platform has been thought to fasten and simplify the prototyping process with the use of a circular platform integrating four rectangular microchips made of thermoplastic. The microchips, dedicated to anion-exchange chromatographic separations, integrate a localized monolithic stationary phase as well as injection and collection reservoirs. The results presented here were obtained with a simplified simulated nuclear spent fuel sample composed of non-radioactive isotopes of Europium and Uranium, in proportion usually found for uranium oxide nuclear spent fuel. While keeping the analytical results consistent with the conventional procedure (extraction yield for Europium of ?97%), the use of the centrifugal microfluidic platform allowed to reduce the volume of liquid needed by a factor of ?250. Thanks to their unique “easy-to-use” features, centrifugal microfluidic platforms are potential successful candidates for the downscaling of chromatographic separation of radioactive samples (automation, multiplexing, easy integration in glove-boxes environment and low cost of maintenance).

Anthony Bruchet; Vélan Taniga; Stéphanie Descroix; Laurent Malaquin; Florence Goutelard; Clarisse Mariet

2013-01-01T23:59:59.000Z

308

"2012 Uranium Marketing Annual Report"  

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

4. Uranium sellers to owners and operators of U.S. civilian nuclear power reactors, 2010-2012" 4. Uranium sellers to owners and operators of U.S. civilian nuclear power reactors, 2010-2012" 2010,2011,2012 "American Fuel Resources, LLC","Advance Uranium Asset Management Ltd. (was Uranium Asset Management)","Advance Uranium Asset Management Ltd. (was Uranium Asset Management)" "AREVA NC, Inc. (was COGEMA, Inc.)","American Fuel Resources, LLC","American Fuel Resources, LLC" "BHP Billiton Olympic Dam Corporation Pty Ltd","AREVA NC, Inc.","AREVA NC, Inc." "CAMECO","BHP Billiton Olympic Dam Corporation Pty Ltd","BHP Billiton Olympic Dam Corporation Pty Ltd" "ConverDyn","CAMECO","CAMECO" "Denison Mines Corp.","ConverDyn","ConverDyn"

309

Thermoacoustic device for nuclear fuel monitoring and heat transfer enhancement  

Science Journals Connector (OSTI)

The Fukushima Dai’ichi nuclear disaster of 2011 exposed the need for self-powered sensors that could transmit the status of the fuel rods within the reactor and in spent fuel ponds that was not dependent upon availability of external electrical power for either sensing or telemetry. One possible solution is the use of a thermoacoustic standing wave engine incorporated within a fuel rod which is heated by the nuclear fuel. The engine’s resonance frequency is correlated to the fuel rod temperature and will be transmitted by sound radiation through the reactor's or storage pond’s surrounding water. In addition to acting as a passive temperature sensor the thermoacoustic device will serve to enhance heat transfer from the fuel to the surrounding heat transfer fluid. When activated the acoustically-driven streaming flow of the gas within the fuel rod will circulate gas away from the nuclear fuel and convectively enhance heat transfer to the surrounding coolant. We will present results for a thermoacousticresonator built into a Nitonic® 60 (stainless steel) fuel rod that can be substituted for conventional fuel rods in the Idaho National Laboratory’s Advanced Test Reactor. This laboratory version is heated electrically. [Work supported by the U.S. Department of Energy.

Randall A. Ali; Steven L. Garrett; James A. Smith; Dale K. Kotter

2012-01-01T23:59:59.000Z

310

Examination of the conversion of the U.S. submarine fleet from highly enriched uranium to low enriched uranium  

E-Print Network [OSTI]

The nuclear reactors used by the U.S. Navy for submarine propulsion are currently fueled by highly enriched uranium (HEU), but HEU brings administrative and political challenges. This issue has been studied by the Navy ...

McCord, Cameron (Cameron Liam)

2014-01-01T23:59:59.000Z

311

Ukraine Fuel Removal: Fact Sheet | National Nuclear Security Administration  

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

Ukraine Fuel Removal: Fact Sheet | National Nuclear Security Administration Ukraine Fuel Removal: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Ukraine Fuel Removal: Fact Sheet Fact Sheet Ukraine Fuel Removal: Fact Sheet Mar 26, 2012 For nearly two decades, the United States and Ukraine have cooperated on a

312

Nuclear Power Generation and Fuel Cycle Report 1997  

Gasoline and Diesel Fuel Update (EIA)

7) 7) Distribution Category UC-950 Nuclear Power Generation and Fuel Cycle Report 1997 September 1997 Energy Information Administration Office of Coal, Nuclear, Electric and Alternate Fuels U.S. Department of Energy Washington, DC 20585 This report was prepared by the Energy Information Administration, the independent statistical and analytical agency within the Department of Energy. The information contained herein should not be construed as advocating or reflecting any policy position of the Department of Energy or of any other organization. Contacts Energy Information Administration/ Nuclear Power Generation and Fuel Cycle Report 1997 ii The Nuclear Power Generation and Fuel Cycle Report is prepared by the U.S. Department of Energy's Energy Information Administration. Questions and comments concerning the contents of the report may be directed to:

313

Public Acceptability of and Preferences for Used Nuclear Fuel...  

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

Acceptance and Preferences for Used Nuclear Fuel Management in the U.S. Hank C. Jenkins-Smith Kuhika Gupta Center for Energy, Security & Society University of Oklahoma Energy...

314

Ukraine Fuel Removal: Fact Sheet | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Ukraine Fuel Removal: Fact Sheet | National Nuclear Security Administration Ukraine Fuel Removal: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Ukraine Fuel Removal: Fact Sheet Fact Sheet Ukraine Fuel Removal: Fact Sheet Mar 26, 2012 For nearly two decades, the United States and Ukraine have cooperated on a

315

Nuclear Power Generation and Fuel Cycle Report 1996  

Gasoline and Diesel Fuel Update (EIA)

6) 6) Distribution Category UC-950 Nuclear Power Generation and Fuel Cycle Report 1996 October 1996 Energy Information Administration Office of Coal, Nuclear, Electric and Alternate Fuels U.S. Department of Energy Washington, DC 20585 This report was prepared by the Energy Information Administration, the independent statistical and analytical agency within the Department of Energy. The information contained herein should not be construed as advocating or reflecting any policy position of the Department of Energy or of any other organization. Energy Information Administration/ Nuclear Power Generation and Fuel Cycle Report 1996 ii Contacts This report was prepared in the Office of Coal, Nuclear, report should be addressed to the following staff Electric and Alternate Fuels by the Analysis and Systems

316

Next-generation nuclear fuel withstands high-temperature accident...  

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

teri.ehresman@inl.gov Bill Cabage (ORNL), 865-574-4399, cabagewh@ornl.gov Next-generation nuclear fuel withstands high-temperature accident conditions IDAHO FALLS - A safer...

317

Plutonium and Reprocessing of Spent Nuclear Fuel  

Science Journals Connector (OSTI)

...might spawn nuclear terrorism. Less than...reprocessing plant. The U.S. nuclear-energy...current fleet of power reactors (15...operational risk of transmutation...future of nuclear power is clarified...constructed plant increased...

Frank N. von Hippel

2001-09-28T23:59:59.000Z

318

Standard specification for uranium oxides with a 235U content of less than 5 % for dissolution prior to conversion to nuclear-grade uranium dioxide  

E-Print Network [OSTI]

1.1 This specification covers uranium oxides, including processed byproducts or scrap material (powder, pellets, or pieces), that are intended for dissolution into uranyl nitrate solution meeting the requirements of Specification C788 prior to conversion into nuclear grade UO2 powder with a 235U content of less than 5 %. This specification defines the impurity and uranium isotope limits for such urania powders that are to be dissolved prior to processing to nuclear grade UO2 as defined in Specification C753. 1.2 This specification provides the nuclear industry with a general standard for such uranium oxide powders. It recognizes the diversity of conversion processes and the processes to which such powders are subsequently to be subjected (for instance, by solvent extraction). It is therefore anticipated that it may be necessary to include supplementary specification limits by agreement between the buyer and seller. 1.3 The scope of this specification does not comprehensively cover all provisions for prevent...

American Society for Testing and Materials. Philadelphia

2005-01-01T23:59:59.000Z

319

Thermal Analysis of Surrogate Simulated Molten Salts with Metal Chloride Impurities for Electrorefining Used Nuclear Fuel  

SciTech Connect (OSTI)

This project is a fundamental study to measure thermal properties (liquidus, solidus, phase transformation, and enthalpy) of molten salt systems of interest to electrorefining operations, which are used in both the fuel cycle research & development mission and the spent fuel treatment mission of the Department of Energy. During electrorefining operations the electrolyte accumulates elements more active than uranium (transuranics, fission products and bond sodium). The accumulation needs to be closely monitored because the thermal properties of the electrolyte will change as the concentration of the impurities increases. During electrorefining (processing techniques used at the Idaho National Laboratory to separate uranium from spent nuclear fuel) it is important for the electrolyte to remain in a homogeneous liquid phase for operational safeguard and criticality reasons. The phase stability of molten salts in an electrorefiner may be adversely affected by the buildup of fission products in the electrolyte. Potential situations that need to be avoided are: (i) build up of fissile elements in the salt approaching the criticality limits specified for the vessel (ii) freezing of the salts due to change in the liquidus temperature and (iii) phase separation (non-homogenous solution) of elements. The stability (and homogeneity) of the phases can potentially be monitored through the thermal characterization of the salts, which can be a function of impurity concentration. This work describes the experimental results of typical salts compositions, consisting of chlorides of strontium, samarium, praseodymium, lanthanum, barium, cerium, cesium, neodymium, sodium and gadolinium (as a surrogate for both uranium and plutonium), used in the processing of used nuclear fuels. Differential scanning calorimetry was used to analyze numerous salt samples providing results on the thermal properties. The property of most interest to pyroprocessing is the liquidus temperature. It was previously known the liquidus temperature of the molten salt would change as spent fuel is processed through the Mk-IV electrorefiner. However, the extent of the increase in liquidus temperature was not known. This work is first of its kind in determining thermodynamic properties of a molten salt electrolyte containing transuranics, fission products and bond sodium. Experimental data concluded that the melting temperature of the electrolyte will become greater than the operating temperature of the Mk-IV ER during current fuel processing campaigns. Collected data also helps predict when the molten salt electrolyte will no longer be able to support electrorefining operations.

Toni Y. Gutknecht; Guy L. Fredrickson; Vivek Utgikar

2012-04-01T23:59:59.000Z

320

Improved Irradiation Performance of Uranium-Molybdenum/Aluminum Dispersion Fuel by Silicon Addition in Aluminum  

SciTech Connect (OSTI)

Uranium-molybdenum fuel particle dispersion in aluminum is a form of fuel under development for conversion of high-power research and test reactors from highly enriched to low-enriched uranium in the U.S. Global Threat Reduction Initiative program (also known as the Reduced Enrichment for Research and Test Reactors program). Extensive irradiation tests have been conducted to find a solution for problems caused by interaction layer growth and pore formation between U-Mo and Al. Adding a small amount of Si (up to [approximately]5 wt%) in the Al matrix was one of the proposed remedies. The effect of silicon addition in the Al matrix was examined using irradiation test results by comparing side-by-side samples with different Si additions. Interaction layer growth was progressively reduced with increasing Si addition to the matrix Al, up to 4.8 wt%. The Si addition also appeared to delay pore formation and growth between the U-Mo and Al.

Yeon Soo Kim; G. L. Hofman; A. B. Robinson; D. M. Wachs

2013-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

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

322

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

SciTech Connect (OSTI)

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

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

2012-05-01T23:59:59.000Z

323

Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF)  

Broader source: Energy.gov [DOE]

GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor...

324

Assessment of the environmental footprint of nuclear energy systems. Comparison between closed and open fuel cycles  

Science Journals Connector (OSTI)

Abstract Energy perspectives for the current century are dominated by the anticipated significant increase of energy needs. Particularly, electricity consumption is anticipated to increase by a factor higher than two before 2050. Energy choices are considered as structuring political choices that implies a long-standing and stable policy based on objective criteria. LCA (life cycle analysis) is a structured basis for deriving relevant indicators which can allow the comparison of a wide range of impacts of different energy sources. Among the energy-mix, nuclear power is anticipated to have very low GHG-emissions. However, its viability is severely addressed by the public opinion after the Fukushima accident. Therefore, a global LCA of the French nuclear fuel cycle was performed as a reference model. Results were compared in terms of impact with other energy sources. It emphasized that the French nuclear energy is one of the less impacting energy, comparable with renewable energy. In a second, part, the French scenario was compared with an equivalent open fuel cycle scenario. It demonstrates that an open fuel cycle would require about 16% more natural uranium, would have a bigger environmental footprint on the “non radioactive indicators” and would produce a higher volume of high level radioactive waste.

Ch. Poinssot; S. Bourg; N. Ouvrier; N. Combernoux; C. Rostaing; M. Vargas-Gonzalez; J. Bruno

2014-01-01T23:59:59.000Z

325

LANL researchers improve path to producing uranium compounds, candidates  

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

Researchers improve path to producing uranium compounds Researchers improve path to producing uranium compounds LANL researchers improve path to producing uranium compounds, candidates for advanced nuclear fuels Enhance the ability to develop advanced nuclear fuels in a safer, simpler manner. April 7, 2011 This illustration shows the structures of UI4(1,4-dioxane)2 (left) and the UI3(1,4-dioxane)1.5 complexes. This illustration shows the structures of UI4(1,4-dioxane)2 (left) and the UI3(1,4-dioxane)1.5 complexes. Contact Kevin Roark Communicatons Office (505) 665-9202 Email LOS ALAMOS, New Mexico, April 7, 2010- Advances made by researchers at Los Alamos National Laboratory could enhance the ability of scientists to develop advanced nuclear fuels in a safer, simpler manner. Uranium chemistry research relies heavily on a variety of uranium "starting

326

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

SciTech Connect (OSTI)

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

Schneider, Erich; Scopatza, Anthony [The University of Texas at Austin, 1 University Station C2200, Austin TX 78712 (United States); Deinert, Mark [The University of Texas at Austin, 1 University Station C2200, Austin TX 78712 (United States); Cornell University, Ithaca NY 14853 (United States)

2007-07-01T23:59:59.000Z

327

Oklo natural nuclear reactor  

Science Journals Connector (OSTI)

Although the possibility of natural was considered in the 1950s, this prediction was not confirmed until 1972, when a routine analysis of a uranium ore standard at the nuclear-fuel processing plant in Pierre...

Mark J. Rigali; Bartholomew Nagy

1998-01-01T23:59:59.000Z

328

FURTHER ASSESSMENTS OF THE ATTRACTIVENESS OF MATERIALS IN ADVANCED NUCLEAR FUEL CYCLES FROM A SAFEGUARDS PERSPECTIVE  

SciTech Connect (OSTI)

This paper summarizes the results of an extension to an earlier study [ ] that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the PUREX, UREX+, and COEX reprocessing schemes. This study focuses on the materials associated with the UREX, COEX, THOREX, and PYROX reprocessing schemes. This study also examines what is required to render plutonium as “unattractive.” Furthermore, combining the results of this study with those from the earlier study permits a comparison of the uranium and thorium based fuel cycles on the basis of the attractiveness of the SNM associated with each fuel cycle. Both studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of “attractiveness levels” that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [ ]. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed.

Bathke, C. G.; Jarvinen, G. D.; Wallace, R. K.; Ireland, J. R.; Johnson, M. W.; Sleaford, Brad W.; Ebbinghaus, B. B.; Bradley, Keith S.; Collins, Brian A.; Smith, Brian W.; Prichard, Andrew W.

2008-10-01T23:59:59.000Z

329

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410  

SciTech Connect (OSTI)

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for use as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)

Phillips, Chris; Willis, William; Carter, Robert [EnergySolutions Federal EPC., 2345 Stevens Drive, Richland, WA, 99354 (United States)] [EnergySolutions Federal EPC., 2345 Stevens Drive, Richland, WA, 99354 (United States); Baker, Stephen [UK National Nuclear Laboratory, Warrington, Cheshire (United Kingdom)] [UK National Nuclear Laboratory, Warrington, Cheshire (United Kingdom)

2013-07-01T23:59:59.000Z

330

International Nuclear Fuel Cycle Fact Book. Revision 5  

SciTech Connect (OSTI)

This Fact Book has been compiled in an effort to provide: (1) an overview of worldwide nuclear power and fuel cycle programs; and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

1985-01-01T23:59:59.000Z

331

International nuclear fuel cycle fact book. Revision 4  

SciTech Connect (OSTI)

This Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids - international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

1984-03-01T23:59:59.000Z

332

MOX fuel arrangement for nuclear core  

DOE Patents [OSTI]

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

2001-07-17T23:59:59.000Z

333

Mox fuel arrangement for nuclear core  

DOE Patents [OSTI]

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

2001-05-15T23:59:59.000Z

334

MOX fuel arrangement for nuclear core  

DOE Patents [OSTI]

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

Kantrowitz, M.L.; Rosenstein, R.G.

1998-10-13T23:59:59.000Z

335

MOX fuel arrangement for nuclear core  

DOE Patents [OSTI]

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

1998-01-01T23:59:59.000Z

336

Depleted uranium  

Science Journals Connector (OSTI)

The potential health effects arising from exposure to depleted uranium have been much in the news of late. Naturally occurring uranium contains the radioisotopes 238U (which dominates, at a current molar proportion of 99.3%), 235U and a small amount of 234U. Depleted uranium has an isotopic concentration of 235U that is below the 0.7% found naturally. This is either because the uranium has passed through a nuclear reactor which uses up some of the fissile 235U that fuels the fission chain-reaction, or because it is the uranium that remains when enriched uranium with an elevated concentration of 235U is produced in an enrichment plant, or because of a combination of these two processes. Depleted uranium has a lower specific activity than naturally occurring uranium because of the lower concentrations of the more radioactive isotopes 235U and 234U, but account must be taken of any contaminating radionuclides or exotic radioisotopes of uranium if the uranium has been irradiated. Uranium is a particularly dense element (about twice as dense as lead), and this property makes it useful in certain military applications, such as armour-piercing munitions. Depleted uranium, rather than natural uranium, is used because of its availability and, since the demise of the fast breeder reactor programme, the lack of alternative use. Depleted uranium weapons were used in the Gulf War of 1990 and also, to a lesser extent, more recently in the Balkans. This has led to speculation that depleted uranium may be associated with `Gulf War Syndrome', or other health effects that have been reported by military and civilian personnel involved in these conflicts and their aftermath. Although, on the basis of present scientific knowledge, it seems most unlikely that exposure to depleted uranium at the levels concerned could produce a detectable excess of adverse health effects, and in such a short timescale, the issue has become one of general concern and contention. As a consequence, any investigation needs to be thorough to produce sufficiently comprehensive evidence to stand up to close scrutiny and gain the support of the public, whatever the conclusions. Unfortunately, it is the nature of such inquiries that they take time, which is frustrating for some. In the UK, the Royal Society has instigated an independent investigation into the health effects of depleted uranium by a working group chaired by Professor Brian Spratt. This inquiry has been underway since the beginning of 2000. The working group's findings will be reviewed by a panel appointed by the Council of the Royal Society, and it is anticipated that the final report will be published in the summer of 2001. Further details can be found at www.royalsoc.ac.uk/templates/press/showpresspage.cfm?file=2001010801.txt. Nick Priest has summarised current knowledge on the toxicity (both radiological and chemical) of depleted uranium in a commentary in The Lancet (27 January 2001, 357 244-6). For those wanting to read a comprehensive review of the literature, in 1999 RAND published `A Review of the Scientific Literature as it Pertains to Gulf War Illnesses, Volume 7: Depleted Uranium' by Naomi Harley and her colleagues, which can be found at www.rand.org/publications/MR/MR1018.7/MR1018.7.html. An interesting article by Jan Olof Snihs and Gustav Akerblom entitled `Use of depleted uranium in military conflicts and possible impact on health and environment' was published in the December 2000 issue of SSI News (pp 1-8), and can be found at the website of the Swedish Radiation Protection Institute: www.ssi.se/tidningar/PDF/lockSSIn/SSI-news2000.pdf. Last year, a paper was published in the June issue of this Journal that is of some relevance to depleted uranium. McGeoghegan and Binks (2000 J. Radiol. Prot. 20 111-37) reported the results of their epidemiological study of the health of workers at the Springfields uranium production facility near Preston during 1946-95. This study included almost 14 000 radiation workers. Although organ-specific doses due to uranium are not yet available for these worker

Richard Wakeford

2001-01-01T23:59:59.000Z

337

Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management  

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

NRC's NRC's Integrated Strategy for NRC s Integrated Strategy for Spent Fuel Management Earl Easton 1 U.S. Nuclear Regulatory Commission May 25, 2010 Road to Yucca Mountain * 20+ years of preparation for the licensing i review * DOE application received in June 2008 and accepted for review in September 2008 * President Obama pursues alternatives to Yucca Mountain * DOE motion to withdraw in March 2010 2 * DOE motion to withdraw in March 2010 * Blue Ribbon Commission on America's Nuclear Future 2 Growing Spent Fuel Inventory Cumulative Used Nuclear Fuel Scenarios 50,000 100,000 150,000 200,000 250,000 Metric Tons 3 - 50,000 2010 2015 2020 2025 2030 2035 2040 2045 2050 Year Reference: Crozat, March 2010 Integrated Strategy * In response to the evolving national debate on spent fuel management strategy, NRC initiated a number of actions:

338

Review of Used Nuclear Fuel Storage and Transportation Technical Gap  

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

Analysis Analysis Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analysis While both wet and dry storage have been shown to be safe options for storing used nuclear fuel (UNF), the focus of the program is on dry storage of commercial UNF at reactor or centralized locations. This report focuses on the knowledge gaps concerning extended storage identified in numerous domestic and international investigations and provides the Used Fuel Disposition Campaign"s (UFDC) gap description, any alternate gap descriptions, the rankings by the various organizations, evaluation of the priority assignment, and UFDC-recommended action based on the comparison. Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analysis More Documents & Publications

339

Spent nuclear fuel discharges from US reactors 1992  

SciTech Connect (OSTI)

This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

Not Available

1994-05-05T23:59:59.000Z

340

Engineering assessment of inactive uranium mill tailings: Phillips/United Nuclear site, Ambrosia Lake, New Mexico  

SciTech Connect (OSTI)

Ford, Bacon and Davis Utah, Inc., has reevaluated the Phillips/United Nuclear site in order to revise the December 1977 engineering assessment of the problems resulting from the existence of radioactive uranium mill tailings at Ambrosia Lake, New Mexico. This engineering assessment has included the preparation of topographic maps, the performance of core drillings and radiometric measurements sufficient to determine areas and volumes of tailings and radiation exposures of individuals and nearby populations, the investigations of site hydrology and meteorology, and the evaluation and costing of alternative corrective actions. Radon gas released from 2.6 million dry tons of tailings at the Phillips/United Nuclear site constitutes the most significant environmental impact, although windblown tailings and external gamma radiation also are factors. The four alternative actions presented in this engineering assessment range from millsite decontamination with the addition of 3 m of stabilization cover material, to removal of the tailings to remote disposal sites and decontamination of the tailings site. Cost estimates for the four options range from about $21,500,000 for stabilization in-place, to about $45,200,000 for disposal at a distance of about 15 mi. Three principal alternatives for the reprocessing of the Phillips/United Nuclear tailings were examined: heap leaching; treatment at an existing mill; and reprocessing at a new conventional mill constructed for tailings reprocessing.The cost of the uranium recovered would be about $87/lb of U/sub 3/O/sub 8/ by either heap leach or conventional plant process. The spot market price for uranium was $25/lb early in 1981. Reprocessing the Phillips/United Nuclear tailings for uranium recovery does not appear to be economically attractive under present or foreseeable market conditions.

none,

1981-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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341

Plutonium and Reprocessing of Spent Nuclear Fuel  

Science Journals Connector (OSTI)

...Repository for the Disposal of Spent Nuclear...Radioactive Waste at Yucca Mountain (YMP-0106...not committed funding to build...Repository for the Disposal of Spent Nuclear...Radioactive Waste at Yucca Mountain (YMP-0106, Yucca Mountain Project, North...

Frank N. von Hippel

2001-09-28T23:59:59.000Z

342

Nuclear energy: Thorium fuel has risks  

Science Journals Connector (OSTI)

... thorium's use in declared nuclear activities, and greater vigilance is needed to protect against surreptitious activities involving this element. ...

Stephen F. Ashley; Geoffrey T. Parks; William J. Nuttall; Colin Boxall; Robin W. Grimes

2012-12-05T23:59:59.000Z

343

An evaluation of thermal modeling techniques utilized for nuclear fuel rods  

E-Print Network [OSTI]

like to thank my graduate advisor, Dr. K. L. Peddicord, for his technical advice and guidance throughout this project and my studies in Nuclear Engineering at Texas AgiM University. Thanks are also extended to Dr. Hassan and Dr. Caton for reviewing.... Burnup Fission Gas Rdease Fuel Tltermal Conductivity Fuel Cracking Fuel Creep Rate Fuel Relocadon Fuel VIrermal Expansion Fuel Rod Tltermal Power Fuel Telltpelanaes Fuel Stress es Fuel Strains Fission Rate Fuel-Cladding Gap Heat...

Simmons, Jeffrey Warren

2012-06-07T23:59:59.000Z

344

Fuel bundle design for enhanced usage of plutonium fuel  

DOE Patents [OSTI]

A nuclear fuel bundle includes a square array of fuel rods each having a concentration of enriched uranium and plutonium. Each rod of an interior array of the rods also has a concentration of gadolinium. The interior array of rods is surrounded by an exterior array of rods void of gadolinium. By this design, usage of plutonium in the nuclear reactor is enhanced.

Reese, Anthony P. (San Jose, CA); Stachowski, Russell E. (Fremont, CA)

1995-01-01T23:59:59.000Z

345

SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION  

SciTech Connect (OSTI)

With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can supply useful information to analysts evaluating spent fuel subcriticality. The original purpose of the subcritical measurements was to validate computer model predictions that spent N Reactor fuel of a particular, typical exposure (2740 MWd/t) had a critical mass equal to twice that of unexposed fuel of the same type. The motivation for performing this work was driven by the need to increase spent fuel storage limits. These subcritical measurements confirmed the computer model predictions.

SCHWINKENDORF, K.N.

2006-05-12T23:59:59.000Z

346

TOWARDS BENCHMARK MEASUREMENTS FOR USED NUCLEAR FUEL ASSAY USING A LEAD SLOWING-DOWN SPECTROMETER  

E-Print Network [OSTI]

for spent fuel testing. The characterization of spent fuel is particularly important for nuclear safeguardsTOWARDS BENCHMARK MEASUREMENTS FOR USED NUCLEAR FUEL ASSAY USING A LEAD SLOWING-DOWN SPECTROMETER B) is considered as a possible option for non- destructive assay of fissile material in used nuclear fuel

Danon, Yaron

347

Software: Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering  

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

Analysis > Analysis > Software Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Overview Current Projects Software Nuclear Plant Dynamics and Safety Nuclear Data Program Advanced Reactor Development Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Reactor Physics and Fuel Cycle Analysis Software Bookmark and Share An extensive powerful suite of computer codes developed and validated by the NE Division and its predecessor divisions at Argonne supports the development of fast reactors; many of these codes are also applicable to other reactor types. A brief description of these codes follows. Contact

348

Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography  

E-Print Network [OSTI]

This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry storage containers. We show that the tracking of individual muons, as they enter and exit a structure, can potentially improve the accuracy and availability of data on nuclear waste and the contents of Dry Storage Containers (DSC) used for spent fuel storage at CANDU plants. This could be achieved in near real time, with the potential for unattended and remotely monitored operations. We show that the expected sensitivity, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy.

Jonkmans, G; Jewett, C; Thompson, M

2012-01-01T23:59:59.000Z

349

A framework and methodology for nuclear fuel cycle transparency.  

SciTech Connect (OSTI)

A key objective to the global deployment of nuclear technology is maintaining transparency among nation-states and international communities. By providing an environment in which to exchange scientific and technological information regarding nuclear technology, the safe and legitimate use of nuclear material and technology can be assured. Many nations are considering closed or multiple-application nuclear fuel cycles and are subsequently developing advanced reactors in an effort to obtain some degree of energy self-sufficiency. Proliferation resistance features that prevent theft or diversion of nuclear material and reduce the likelihood of diversion from the civilian nuclear power fuel cycle are critical for a global nuclear future. IAEA Safeguards have been effective in minimizing opportunities for diversion; however, recent changes in the global political climate suggest implementation of additional technology and methods to ensure the prompt detection of proliferation. For a variety of reasons, nuclear facilities are becoming increasingly automated and will require minimum manual operation. This trend provides an opportunity to utilize the abundance of process information for monitoring proliferation risk, especially in future facilities. A framework that monitors process information continuously can lead to greater transparency of nuclear fuel cycle activities and can demonstrate the ability to resist proliferation associated with these activities. Additionally, a framework designed to monitor processes will ensure the legitimate use of nuclear material. This report describes recent efforts to develop a methodology capable of assessing proliferation risk in support of overall plant transparency. The framework may be tested at the candidate site located in Japan: the Fuel Handling Training Model designed for the Monju Fast Reactor at the International Cooperation and Development Training Center of the Japan Atomic Energy Agency.

McClellan, Yvonne; York, David L.; Inoue, Naoko (Japan Atomic Energy Agency, Ibaraki, Japan); Love, Tracia L.; Rochau, Gary Eugene

2006-02-01T23:59:59.000Z

350

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321  

SciTech Connect (OSTI)

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States); Moore, E.N. [Moore Nuclear Energy, LLC (United States)] [Moore Nuclear Energy, LLC (United States)

2013-07-01T23:59:59.000Z

351

Towards a desalination initiative using cogeneration with an advanced reactor type and uranium recovered from Moroccan phosphoric acid production  

Science Journals Connector (OSTI)

Morocco is known to be among the first few countries to produce phosphate and phosphoric acid. Moroccan phosphate contains substantial amounts of uranium. This uranium can be recovered from the phosphate ore as a by-product during the production of phosphoric acid. Uranium extraction processes linked with phosphoric acid fabrication have been used industrially in some countries. This is done mainly by solvent extraction. Although, the present price of uranium is low in the international market, such uranium recovery could be considered as a side product of phosphoric acid production. The price of uranium has a very small impact on the cost of nuclear energy obtained from it. This paper focuses on the extraction of uranium salt from phosphate rock. If uranium is recovered in Morocco in the proposed manner, it could serve as feed for a number of nuclear power plants. The natural uranium product would have to be either enriched or blended as mixed-oxide fuel to manufacture adequate nuclear fuel. Part of this fuel would feed a desalination initiative using a high temperature reactor of the new generation, chosen for its intrinsic safety, sturdiness, ease of maintenance, thermodynamic characteristics and long fuel life between reloads, that is, good economy. ?n international cooperation based on commercial contract schemes would concern: the general project and uranium extraction; uranium enrichment and fuel fabrication services; the nuclear power plant; and the desalination plant. This paper presents the overall feasibility of the general project with some quantitative preliminary figures and cost estimates.

Michel Lung; Abdelaali Kossir; Driss Msatef

2005-01-01T23:59:59.000Z

352

Integrated data base report--1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect (OSTI)

The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and commercial and U.S. government-owned radioactive wastes. Inventories of most of these materials are reported as of the end of fiscal year (FY) 1996, which is September 30, 1996. Commercial SNF and commercial uranium mill tailings inventories are reported on an end-of-calendar year (CY) basis. All SNF and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are SNF, high-level waste, transuranic waste, low-level waste, uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, naturally occurring and accelerator-produced radioactive material, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through FY 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

NONE

1997-12-01T23:59:59.000Z

353

Radioactive Flow Characterization for Real-Time Detection Systems in UREX+ Nuclear Fuel Reprocessing  

E-Print Network [OSTI]

plants. Current large-scale commercial reprocessing facilities around the world make use of the PUREX (Plutonium-Uranium Extraction) method, which separates out a pure plutonium and pure uranium stream and puts all the remaining isotopes... products in the fuel. The leading technology option in the U.S. for reprocessing is a sequence of processing methods known as UREX+ (Uranium Extraction+). However, an industrial scale facility implementing this separation procedure will require...

Hogelin, Thomas Russell

2011-02-22T23:59:59.000Z

354

International nuclear fuel cycle fact book. [Contains glossary  

SciTech Connect (OSTI)

As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is a consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

Leigh, I.W.; Lakey, L.T.; Schneider, K.J.; Silviera, D.J.

1987-01-01T23:59:59.000Z

355

International Nuclear Fuel Cycle Fact Book. Revision 12  

SciTech Connect (OSTI)

As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need exists costs for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book has been compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NMEA activities reports; and proceedings of conferences and workshops. The data listed typically do not reflect any single source but frequently represent a consolidation/combination of information.

Leigh, I.W.

1992-05-01T23:59:59.000Z

356

An overview of measurement methods for special nuclear material in spent nuclear fuel  

SciTech Connect (OSTI)

Summary results from a survey of nondestructive assay measurement methods applicable to the measurement of the special nuclear material content of spent nuclear fuel are described. The role of nuclear materials measurements in the domestic and international safeguarding of spent nuclear fuel in the United States' federal waste management system has yet to be determined. An understanding of the characteristics and capabilities of the potentially applicable measurement systems should provide valuable information to the developers of the safeguards approaches for the monitored retrievable storage and final disposal systems. The discussion focuses on the general characteristics of the identified direct and indirect measurement methods. 3 refs., 1 tab.

Moran, B.W.; Reich, W.J.

1989-07-01T23:59:59.000Z

357

Double-clad nuclear-fuel safety rod  

DOE Patents [OSTI]

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

McCarthy, W.H.; Atcheson, D.B.

1981-12-30T23:59:59.000Z

358

Apparatus for injection casting metallic nuclear energy fuel rods  

DOE Patents [OSTI]

Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

Seidel, Bobby R. (Idaho Falls, ID); Tracy, Donald B. (Firth, ID); Griffiths, Vernon (Butte, MT)

1991-01-01T23:59:59.000Z

359

Concept of development of nuclear power based on LMFBR operation in open nuclear fuel cycle  

SciTech Connect (OSTI)

The preliminary assessments performed show that it is reasonable to investigate in the future the possibilities of FBR efficient operation with the open NFC. To improve its safety it is expedient to use the lead-bismuth alloy as a coolant. In order to operate with depleted uranium make-up it is necessary to meet a number of requirements providing the reactor criticality due to plutonium build-up and BR > 1. These requirements are as follows: a large core (20--25 m{sup 3}); a high fuel volume fraction (> 60%); utilization of dense metallic fuel; a high fuel burn-up--at a level of 20% of h.a. Making use of these reactors should allow the NP fuel base to be extended more than 10 times without making NFC closed. It provides improving NP safety during a sufficiently long stage of its development.

Toshinsky, G.I. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation)

1996-08-01T23:59:59.000Z

360

Synergistic smart fuel for in-pile nuclear reactor measurements  

SciTech Connect (OSTI)

The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

2013-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Quantitative assessment of proposals on assurance of nuclear fuel supply  

SciTech Connect (OSTI)

The assurance of nuclear fuel supply has the potential to contribute to balancing peaceful use of nuclear power and nuclear nonproliferation. 5 proposals which provide the backup supply of the enrichment service in case of supply disruption, are investigated in this study. We investigated the 20 NPT countries which are non-nuclear-weapon states and possess operable commercial LWRs in October 2012 as potential participants for each proposal. As a result of literature researching, we have extracted factors that can be considered as important for a country to participate or not participate in the assurance of nuclear fuel supply. Then we have computed incentive and disincentive parameters for each country. The results show that the participation expectancy decreases in the order of IAEA Fuel Bank proposal, Russian LEU Reserve proposal, AFS proposal, WNA proposal and 6-Country proposal. The 'IAEA fuel bank proposal' would be triggered in case of the supply disruption which cannot be solved by the market mechanism and bilateral agreements.

Tanaka, T.; Kuno, Y.; Tanaka, S. [University of Tokyo, 7-3-1 Hongou, Bunkyou-ku, Tokyo 112-0005 (Japan)

2013-07-01T23:59:59.000Z

362

Environmental Impact of the Nuclear Fuel Cycle: Fate of Actinides  

SciTech Connect (OSTI)

The resurgence of nuclear power as a strategy for reducing greenhouse gas (GHG) emissions has, in parallel, revived interest in the environmental impact of actinides. Just as GHG emissions are the main environmental impact of the combustion of fossil fuels, the fate of actinides, consumed and produced by nuclear reactions, determines whether nuclear power is viewed as an environmentally “friendly” source of energy. In this article, we summarize the sources of actinides in the nuclear fuel cycle, how actinides are separated by chemical processing, the development of actinide-bearing materials, and the behavior of actinides in the environment. At each stage, actinides present a unique and complicated behavior because of the 5f electronic configurations.

Ewing, Rodney C.; Runde, W.; Albrecht-Schmitt, Thomas E.

2010-01-01T23:59:59.000Z

363

Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications  

SciTech Connect (OSTI)

The feasibility of the fabrication of tungsten based nuclear fuel cermets via Spark Plasma Sintering (SPS) is investigated in this work. CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar properties of these materials. This study shows that after a short time sintering, greater than 90 % density can be achieved, which is suitable to possess good strength as well as the ability to contain fission products. The mechanical properties and the densities of the samples are also investigated as functions of the applied pressures during the sintering.

Yang Zhong; Robert C. O'Brien; Steven D. Howe; Nathan D. Jerred; Kristopher Schwinn; Laura Sudderth; Joshua Hundley

2011-11-01T23:59:59.000Z

364

Characterization of Nuclear Fuel using Multivariate Statistical Analysis  

SciTech Connect (OSTI)

Various combinations of reactor type and fuel composition have been characterized using principle components analysis (PCA) of the concentrations of 9 U and Pu isotopes in the 10 fuel as a function of burnup. The use of PCA allows the reduction of the 9-dimensional data (isotopic concentrations) into a 3-dimensional approximation, giving a visual representation of the changes in nuclear fuel composition with burnup. Real-world variation in the concentrations of {sup 234}U and {sup 236}U in the fresh (unirradiated) fuel was accounted for. The effects of reprocessing were also simulated. The results suggest that, 15 even after reprocessing, Pu isotopes can be used to determine both the type of reactor and the initial fuel composition with good discrimination. Finally, partial least squares discriminant analysis (PSLDA) was investigated as a substitute for PCA. Our results suggest that PLSDA is a better tool for this application where separation between known classes is most important.

Robel, M; Robel, M; Robel, M; Kristo, M J; Kristo, M J

2007-11-27T23:59:59.000Z

365

Sensitivity analysis and optimization of the nuclear fuel cycle  

SciTech Connect (OSTI)

A sensitivity study has been conducted to assess the robustness of the conclusions presented in the MIT Fuel Cycle Study. The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycles. The options include limited recycling in LWRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. The analysis allowed optimization of the fast reactor conversion ratio with respect to desired fuel cycle performance characteristics. The following parameters were found to significantly affect the performance of recycling technologies and their penetration over time: Capacity Factors of the fuel cycle facilities, Spent Fuel Cooling Time, Thermal Reprocessing Introduction Date, and in core and Out-of-core TRU Inventory Requirements for recycling technology. An optimization scheme of the nuclear fuel cycle is proposed. Optimization criteria and metrics of interest for different stakeholders in the fuel cycle (economics, waste management, environmental impact, etc.) are utilized for two different optimization techniques (linear and stochastic). Preliminary results covering single and multi-variable and single and multi-objective optimization demonstrate the viability of the optimization scheme. (authors)

Passerini, S.; Kazimi, M. S.; Shwageraus, E. [Massachusetts Inst. of Technology, Dept. of Nuclear Science and Engineering, 77 Massachusetts Avenue, Cambridge, MA 02138 (United States)

2012-07-01T23:59:59.000Z

366

The Need for Confirmatory Experiments on the Radioactive Source Term from Potential Sabotage of Spent Nuclear Fuel Casks  

SciTech Connect (OSTI)

A technical review is presented of experiment activities and state of knowledge on air-borne, radiation source terms resulting from explosive sabotage attacks on spent reactor fuel subassemblies in shielded casks. Current assumptions about the behavior of irradiated fuel are largely based on a limited number of experimental results involving unirradiated, depleted uranium dioxide ''surrogate'' fuel. The behavior of irradiated nuclear fuel subjected to explosive conditions could be different from the behavior of the surrogate fuel, depending on the assumptions made by the evaluator. Available data indicate that these potential differences could result in errors, and possible orders-of-magnitude overestimates of aerosol dispersion and potential health effects from sabotage attacks. Furthermore, it is suggested that the current assumptions used in arriving at existing regulations for the transportation and storage of spent fuel in the U.S. are overly conservative. This, in turn, has led to potentially higher-than-needed operating expenses for those activities. A confirmatory experimental program is needed to develop a realistic correlation between source terms of irradiated fuel and unirradiated fuel. The motivations for performing the confirmatory experimental program are also presented.

PHILBIN, JEFFREY S.; HOOVER, MARK D.; NEWTON, GEORGE J.

2002-04-01T23:59:59.000Z

367

Nuclear Fuel Cycle Reasoner: PNNL FY13 Report  

SciTech Connect (OSTI)

In Fiscal Year 2012 (FY12) PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In Fiscal Year 2013 (FY13) the SNAP demonstration was enhanced with respect to query and navigation usability issues.

Hohimer, Ryan E.; Strasburg, Jana D.

2013-09-30T23:59:59.000Z

368

Non-proliferation: A nuclear exchange  

Science Journals Connector (OSTI)

... the only visible hint of what once lay inside the building: enough highly enriched uranium reactor fuel to make more than 18 nuclear bombs. A group of visiting journalists and ... the fuel's uranium will be converted into a safer form for use in power reactors. ...

Geoff Brumfiel

2010-10-11T23:59:59.000Z

369

E-Print Network 3.0 - alternative nuclear fuel Sample Search...  

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

Summary: of electricity from nuclear power plants is far less than any of the alternative energy technologies now contem... Processing of Nuclear Fuel, EGRN 430 ...

370

Design of Nanosensors for Fissile Materials in Nuclear Waste Water  

Science Journals Connector (OSTI)

Uranium and plutonium ions and their complexes play an important role in nuclear fuel reprocessing, and their trace characterization is important in nuclear forensics. In this work, we perform ab initio density functional theory calculations of U and Pu ...

Narendra Kumar; Jorge M. Seminario

2013-10-16T23:59:59.000Z

371

Development of a nuclear fuel cycle transparency framework.  

SciTech Connect (OSTI)

Nuclear fuel cycle transparency can be defined as a confidence building approach among political entities to ensure civilian nuclear facilities are not being used for the development of nuclear weapons. Transparency concepts facilitate the transfer of nuclear technology, as the current international political climate indicates a need for increased methods of assuring non-proliferation. This research develops a system which will augment current non-proliferation assessment activities undertaken by U.S. and international regulatory agencies. It will support the export of nuclear technologies, as well as the design and construction of Gen. IV energy systems. Additionally, the framework developed by this research will provide feedback to cooperating parties, thus ensuring full transparency of a nuclear fuel cycle. As fuel handling activities become increasingly automated, proliferation or diversion potential of nuclear material still needs to be assessed. However, with increased automation, there exists a vast amount of process data to be monitored. By designing a system that monitors process data continuously, and compares this data to declared process information and plant designs, a faster and more efficient assessment of proliferation risk can be made. Figure 1 provides an illustration of the transparency framework that has been developed. As shown in the figure, real-time process data is collected at the fuel cycle facility; a reactor, a fabrication plant, or a recycle facility, etc. Data is sent to the monitoring organization and is assessed for proliferation risk. Analysis and recommendations are made to cooperating parties, and feedback is provided to the facility. The analysis of proliferation risk is based on the following factors: (1) Material attractiveness: the quantification of factors relevant to the proliferation risk of a certain material (e.g., highly enriched Pu-239 is more attractive than that of lower enrichment) (2) The static (baseline) risk: the quantification of risk factors regarding the expected value of proliferation risk under normal (not proliferating) operations. (3) The dynamic (changing) risk: the quantification of risk factors regarding the observed value of proliferation risk, based on monitor signals from facility operations. This framework could be implemented at facilities which have been exported (for instance, to third world countries), or facilities located in sensitive countries. Sandia National Laboratories is currently working with the Japan Nuclear Cycle Development Institute (JNC) to implement a demonstration of nuclear fuel cycle transparency technology at the Fuel Handling Training Model designed for the Monju Fast Reactor at the International Cooperation and Development Training Center in Japan. This technology has broad applications, both in the U.S. and abroad. Following the demonstration, we expect to begin further testing of the technology at an Enrichment Facility, a Fast Reactor, and at a Recycle Facility.

Love, Tracia L.

2005-04-01T23:59:59.000Z

372

MANAGEMENT OF SPENT NUCLEAR FUEL IN FINLAND: POLICY, PAST AND PRESENT PRACTICES, PLANS FOR THE FUTURE  

Science Journals Connector (OSTI)

In Finland, about 1700 tU of spent nuclear fuel has arisen from the operation of the four nuclear power units which were commissioned in late ... 1980’s. Initially the spent fuel management policy was based on se...

E. RUOKOLA

2006-01-01T23:59:59.000Z

373

E-Print Network 3.0 - apex nuclear fuel Sample Search Results  

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

nuclear fuel Search Powered by Explorit Topic List Advanced Search Sample search results for: apex nuclear fuel Page: << < 1 2 3 4 5 > >> 1 FY 2010 Highlights Faculty and Summary:...

374

Locations of Spent Nuclear Fuel and High-Level Radioactive Waste...  

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

Locations of Spent Nuclear Fuel and High-Level Radioactive Waste Locations of Spent Nuclear Fuel and High-Level Radioactive Waste Map of the United States of America showing the...

375

Plasma processing of spent nuclear fuel by two-frequency ion cyclotron resonance heating  

Science Journals Connector (OSTI)

A previously developed method for analyzing the plasma processing of spent nuclear fuel is generalized to a plasma containing multicharged fuel ions. In such a plasma, ion cyclotron resonance heating of nuclear a...

A. V. Timofeev

2009-11-01T23:59:59.000Z

376

Standard guide for drying behavior of spent nuclear fuel  

E-Print Network [OSTI]

1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate d...

American Society for Testing and Materials. Philadelphia

2008-01-01T23:59:59.000Z

377

Changing Biomass, Fossil, and Nuclear Fuel Cycles for Sustainability  

SciTech Connect (OSTI)

The energy and chemical industries face two great sustainability challenges: the need to avoid climate change and the need to replace crude oil as the basis of our transport and chemical industries. These challenges can be met by changing and synergistically combining the fossil, biomass, and nuclear fuel cycles.

Forsberg, Charles W [ORNL

2007-01-01T23:59:59.000Z

378

Method of controlling crystallite size in nuclear-reactor fuels  

DOE Patents [OSTI]

Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

Lloyd, Milton H. (Oak Ridge, TN); Collins, Jack L. (Knoxville, TN); Shell, Sam E. (Oak Ridge, TN)

1985-01-01T23:59:59.000Z

379

Methods and apparatuses for the development of microstructured nuclear fuels  

DOE Patents [OSTI]

Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

Jarvinen, Gordon D. (Los Alamos, NM); Carroll, David W. (Los Alamos, NM); Devlin, David J. (Santa Fe, NM)

2009-04-21T23:59:59.000Z

380

The Overlooked Back End of the Nuclear Fuel Cycle  

Science Journals Connector (OSTI)

...long-term plan for the disposal of nuclear waste...in new fuel and disposal of the subsequent...geologic repository at Yucca Mountain, Nevada, but the...repository for final disposal. To establish...constant source of funding is required to...

Allison M. Macfarlane

2011-09-02T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" 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

Nuclear Fuel Cycle Reasoner: PNNL FY12 Report  

SciTech Connect (OSTI)

Building on previous internal investments and leveraging ongoing advancements in semantic technologies, PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In developing this proof of concept prototype, the utility and relevancy of semantic technologies to the Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D) has been better understood.

Hohimer, Ryan E.; Pomiak, Yekaterina G.; Neorr, Peter A.; Gastelum, Zoe N.; Strasburg, Jana D.

2013-05-03T23:59:59.000Z

382

Dry storage of spent nuclear fuel in UAE – Economic aspect  

Science Journals Connector (OSTI)

Abstract Cost analysis of dry storage of spent nuclear fuel (SNF) discharged from Barakah nuclear power plants in the UAE was performed using three variables: average fuel discharge rate (FD), discount rate (d), and cooling time in a spent fuel pool (Tcool). The costs of dry storage as an interim spent fuel storage option in the UAE were estimated and compared between the following two scenarios: Scenario 1 is ‘accelerated transfer of spent fuel to dry storage’ that SNF will be transferred to dry storage facilities as soon as spent fuel has been sufficiently cooled down in a pool for the dry storage; Scenario 2 is defined as ‘maximum use of spent fuel pool’ that SNF will be stored in a pool as long as possible till the amount of stored SNF in the pool reaches the capacity of the pools and, then, to be moved to dry storage. A sensitivity analysis on the costs was performed and multiple regression analysis was applied to the resulting net present values (NPVs) for Scenarios 1 and 2 and ?NPV that is difference in the net present values between the two scenarios. The results showed that \\{NPVs\\} and ?NPV could be approximately expressed by single equations with the three variables. Among the three variables, the discount rate had the largest effect on the \\{NPVs\\} of the dry storage costs. However, ?NPV was turned out to be equally sensitive to the discount rate and cooling period. Over the ranges of the variables, the additional cost for accelerated fuel transfer (Scenario 1) ranged from 86.4 to 212.9 million $. Calculated using the maximum difference (212.9 M$) between the two scenarios, the accelerated fuel transfer to dry storage could incur the additional electricity rate 8.0 × 10?5 USD/kWh, which is not considered to be significant, compared to the overall electricity generation cost.

Sara Al Saadi; Yongsun Yi

2015-01-01T23:59:59.000Z

383

Survey of Dynamic Simulation Programs for Nuclear Fuel Reprocessing  

SciTech Connect (OSTI)

The absence of any industrial scale nuclear fuel reprocessing in the U.S. has precluded the necessary driver for developing the advanced simulation capability now prevalent in so many other industries. Modeling programs to simulate the dynamic behavior of nuclear fuel separations and processing were originally developed to support the US government’s mission of weapons production and defense fuel recovery. Consequently there has been little effort is the US devoted towards improving this specific process simulation capability during the last two or three decades. More recent work has been focused on elucidating chemical thermodynamics and developing better models of predicting equilibrium in actinide solvent extraction systems. These equilibrium models have been used to augment flowsheet development and testing primarily at laboratory scales. The development of more robust and complete process models has not kept pace with the vast improvements in computational power and user interface and is significantly behind simulation capability in other chemical processing and separation fields.

Troy J. Tranter; Daryl R. Haefner

2008-06-01T23:59:59.000Z

384

Ceramic-composite waste forms from the electrometallurgical treatment of spent nuclear fuel  

Science Journals Connector (OSTI)

Argonne National Laboratory is developing a method to treat spent nuclear fuel in a molten-salt electrorefiner. Glass...

C. Pereira; M. Hash; M. Lewis; M. Richmann

1997-07-01T23:59:59.000Z

385

LIBS Spectral Data for a Mixed Actinide Fuel Pellet Containing Uranium, Plutonium, Neptunium and Americium  

SciTech Connect (OSTI)

Laser-induced breakdown spectroscopy (LIBS) was used to analyze a mixed actinide fuel pellet containing 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2}. The preliminary data shown here is the first report of LIBS analysis of a mixed actinide fuel pellet, to the authors knowledge. The LIBS spectral data was acquired in a plutonium facility at Los Alamos National Laboratory where the sample was contained within a glove box. The initial installation of the glove box was not intended for complete ultraviolet (UV), visible (VIS) and near infrared (NIR) transmission, therefore the LIBS spectrum is truncated in the UV and NIR regions due to the optical transmission of the window port and filters that were installed. The optical collection of the emission from the LIBS plasma will be optimized in the future. However, the preliminary LIBS data acquired is worth reporting due to the uniqueness of the sample and spectral data. The analysis of several actinides in the presence of each other is an important feature of this analysis since traditional methods must chemically separate uranium, plutonium, neptunium, and americium prior to analysis. Due to the historic nature of the sample fuel pellet analyzed, the provided sample composition of 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2} cannot be confirm without further analytical processing. Uranium, plutonium, and americium emission lines were abundant and easily assigned while neptunium was more difficult to identify. There may be several reasons for this observation, other than knowing the exact sample composition of the fuel pellet. First, the atomic emission wavelength resources for neptunium are limited and such techniques as hollow cathode discharge lamp have different dynamics than the plasma used in LIBS which results in different emission spectra. Secondly, due to the complex sample of four actinide elements, which all have very dense electronic energy levels, there may be reactions and interactions occurring within the plasma, such as collisional energy transfer, that might be a factor in the reduction in neptunium emission lines. Neptunium has to be analyzed alone using LIBS to further understand the dynamics that may be occurring in the plasma of the mixed actinide fuel pellet sample. The LIBS data suggests that the emission spectrum for the mixed actinide fuel pellet is not simply the sum of the emission spectra of the pure samples but is dependent on the species present in the plasma and the interactions and reactions that occur within the plasma. Finally, many of the neptunium lines are in the near infrared region which is drastically reduced in intensity by the current optical setup and possibly the sensitivity of the emission detector in the spectral region. Once the optics are replaced and the optical collection system is modified and optimized, the probability of observing emission lines for neptunium might be increased significantly. The mixed actinide fuel pellet was analyzed under the experimental conditions listed in Table 1. The LIBS spectra of the fuel pellet are shown in Figures 1-49. The spectra are labeled with the observed wavelength and atomic species (both neutral (I) and ionic (II)). Table 2 is a complete list of the observed and literature based emission wavelengths. The literature wavelengths have references including NIST Atomic Spectra Database (NIST), B.A. Palmer et al. 'An Atlas of Uranium Emission Intensities in a Hollow Cathode Discharge' taken at the Kitt Peak National Observatory (KPNO), R.L. Kurucz 1995 Atomic Line Data from the Smithsonian Astrophysical Observatory (SAO), J. Blaise et al. 'The Atomic Spectrum of Plutonium' from Argonne National Laboratory (BFG), and M. Fred and F.S. Tomkins, 'Preliminary Term Analysis of Am I and Am II Spectra' (FT). The dash (-) shown under Ionic State indicates that the ionic state of the transition was not available. In the spectra, the dash (-) is replaced with a question mark (?). Peaks that are not assigned are most likely real features and not noise but cannot be confidently assi

Judge, Elizabeth J. [Los Alamos National Laboratory; Berg, John M. [Los Alamos National Laboratory; Le, Loan A. [Los Alamos National Laboratory; Lopez, Leon N. [Los Alamos National Laboratory; Barefield, James E. [Los Alamos National Laboratory

2012-06-18T23:59:59.000Z

386

Nuclear Fuel in a Reactor Accident  

Science Journals Connector (OSTI)

...Three Mile Island: A report to the commissioners and to the public” (Nuclear Regulatory Commission, Washington, DC, 1980). 5...Podcast The contents of this podcast interview represent the opinion of the author and may go beyond the content of the published...

Peter C. Burns; Rodney C. Ewing; Alexandra Navrotsky

2012-03-09T23:59:59.000Z

387

SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION  

SciTech Connect (OSTI)

With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel and four (4) spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data, such as the uncertainty in fuel exposure impact on reactivity and the pulse neutron data evaluation methodology, failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements supply useful information to analysts evaluating spent fuel subcriticality. The original purpose of the subcritical measurements was to validate computer model predictions that spent N Reactor fuel of a particular, typical exposure (2740 MWd/t) had a critical mass equal to twice that of unexposed fuel of the same type. The motivation for performing this work was driven by the need to increase spent fuel storage limits. These subcritical measurements confirmed the computer model predictions.

TOFFER, H.

2006-07-18T23:59:59.000Z

388

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

E-Print Network [OSTI]

Uranium present in low concentration in ocean water has the potential to greatly augment the current fuel reserve for nuclear power generation, but the challenge of extracting it economically remains. Two new designs of ...

Wu, You

389

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

390

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

391

Gamma Ray Mirrors for Direct Measurement of Spent Nuclear Fuel  

SciTech Connect (OSTI)

Direct measurement of the amount of Pu and U in spent nuclear fuel represents a challenge for the safeguards community. Ideally, the characteristic gamma-ray emission lines from different isotopes provide an observable suitable for this task. However, these lines are generally lost in the fierce flux of radiation emitted by the fuel. The rates are so high that detector dead times limit measurements to only very small solid angles of the fuel. Only through the use of carefully designed view ports and long dwell times are such measurements possible. Recent advances in multilayer grazing-incidence gamma-ray optics provide one possible means of overcoming this difficulty. With a proper optical and coating design, such optics can serve as a notch filter, passing only narrow regions of the overall spectrum to a fully shielded detector that does not view the spent fuel directly. We report on the design of a mirror system and a number of experimental measurements.

Pivovaroff, Dr. Michael J. [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL); Ziock, Klaus-Peter [ORNL] [ORNL; Harrison, Mark J [ORNL] [ORNL; Soufli, Regina [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL)

2014-01-01T23:59:59.000Z

392

Verifiable Fuel Cycle Simulation Model (VISION): A Tool for Analyzing Nuclear Fuel Cycle Futures  

SciTech Connect (OSTI)

The nuclear fuel cycle consists of a set of complex components that are intended to work together. To support the nuclear renaissance, it is necessary to understand the impacts of changes and timing of events in any part of the fuel cycle system such as how the system would respond to each technological change, a series of which moves the fuel cycle from where it is to a postulated future state. The system analysis working group of the United States research program on advanced fuel cycles (formerly called the Advanced Fuel Cycle Initiative) is developing a dynamic simulation model, VISION, to capture the relationships, timing, and changes in and among the fuel cycle components to help develop an understanding of how the overall fuel cycle works. This paper is an overview of the philosophy and development strategy behind VISION. The paper includes some descriptions of the model components and some examples of how to use VISION. For example, VISION users can now change yearly the selection of separation or reactor technologies, the performance characteristics of those technologies, and/or the routing of material among separation and reactor types - with the model still operating on a PC in <5 min.

Jacob J. Jacobson; Steven J. Piet; Gretchen E. Matthern; David E. Shropshire; Robert F. Jeffers; A. M. Yacout; Tyler Schweitzer

2010-11-01T23:59:59.000Z

393

Gap Analysis to Support Extended Storage of Used Nuclear Fuel | Department  

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

Gap Analysis to Support Extended Storage of Used Nuclear Fuel Gap Analysis to Support Extended Storage of Used Nuclear Fuel Gap Analysis to Support Extended Storage of Used Nuclear Fuel The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC are responsible for addressing issues regarding the

394

A micro hot test of the Chalmers-GANEX extraction system on used nuclear fuel  

SciTech Connect (OSTI)

In the present study, a 'micro hot test' has been performed using the Chalmers-GANEX (Group Actinide Extraction) system for partitioning of used nuclear fuel. The test included a pre-extraction step using N,N-di-2- ethylhexyl-butyramide (DEHBA) in n-octanol to remove the bulk part of the uranium. This pre-extraction was followed by a group extraction of actinides using the mixture of TBP and CyMe{sub 4}-BTBP in cyclohexanone as suggested in the Chalmers-GANEX process, and a three stage stripping of the extracted actinides. Distribution ratios for the extractions and stripping were determined based on a combination of ?- and ?-spectrometry, as well as ICP-MS measurements. Successful extraction of uranium, plutonium and the minor actinides neptunium, americium and curium was achieved. However, measurements also indicated that co-extraction of europium occurs to some extent during the separation. These results were expected based on previous experiments using trace concentrations of actinides and lanthanides. Since this test was only performed in one stage with respect to the group actinide extraction, it is expected that multi stage tests will give even better results. (authors)

Bauhn, L.; Hedberg, M.; Aneheim, E.; Ekberg, C.; Loefstroem-Engdahl, E.; Skarnemark, G. [Department of Chemical and Biological Engineering, Nuclear Chemistry, Chalmers University of Technology, Kemivaegen 4, SE-412 96 Goeteborg (Sweden)

2013-07-01T23:59:59.000Z

395

Analyzing Nuclear Fuel Cycles from Isotopic Ratios of Waste Products Applicable to Measurement by Accelerator Mass Spectrometry  

SciTech Connect (OSTI)

An extensive study was conducted to determine isotopic ratios of nuclides in spent fuel that may be utilized to reveal historical characteristics of a nuclear reactor cycle. This forensic information is important to determine the origin of unknown nuclear waste. The distribution of isotopes in waste products provides information about a nuclear fuel cycle, even when the isotopes of uranium and plutonium are removed through chemical processing. Several different reactor cycles of the PWR, BWR, CANDU, and LMFBR were simulated for this work with the ORIGEN-ARP and ORIGEN 2.2 codes. The spent fuel nuclide concentrations of these reactors were analyzed to find the most informative isotopic ratios indicative of irradiation cycle length and reactor design. Special focus was given to long-lived and stable fission products that would be present many years after their creation. For such nuclides, mass spectrometry analysis methods often have better detection limits than classic gamma-ray spectroscopy. The isotopic ratios {sup 151}Sm/{sup 146}Sm, {sup 149}Sm/{sup 146}Sm, and {sup 244}Cm/{sup 246}Cm were found to be good indicators of fuel cycle length and are well suited for analysis by accelerator mass spectroscopy.

Biegalski, S R; Whitney, S M; Buchholz, B

2005-08-24T23:59:59.000Z

396

Strategy for the Management and Disposal of Used Nuclear Fuel and  

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

Strategy for the Management and Disposal of Used Nuclear Fuel and Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste The Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste is a framework for moving toward a sustainable program to deploy an integrated system capable of transporting, storing, and disposing of used nuclear fuel and high-level radioactive waste from civilian nuclear power generation, defense, national security and other activities. Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste More Documents & Publications Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste

397

Strategy for the Management and Disposal of Used Nuclear Fuel and  

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

Strategy for the Management and Disposal of Used Nuclear Fuel and Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste Issued on January 11, 2013, the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste is a framework for moving toward a sustainable program to deploy an integrated system capable of transporting, storing, and disposing of used nuclear fuel and high-level radioactive waste from civilian nuclear power generation, defense, national security and other activities. Strategy for the Management and Disposal of Used Nuclear Fuel and High Level Radioactive Waste.pdf More Documents & Publications Strategy for the Management and Disposal of Used Nuclear Fuel and

398

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect (OSTI)

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

A. Alsaed

2005-07-28T23:59:59.000Z

399

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect (OSTI)

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

J.K. Knudson

2003-10-02T23:59:59.000Z

400

Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input  

SciTech Connect (OSTI)

A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

Monado, Fiber, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

2014-02-12T23:59:59.000Z

Note: This page contains sample records for the topic "uranium fuel nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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401

Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results  

SciTech Connect (OSTI)

This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

Rechard, R.P. [ed.

1995-03-01T23:59:59.000Z

402

Mechanical Properties of Nuclear Fuel Surrogates using Picosecond Laser Ultrasonics  

SciTech Connect (OSTI)

Detailed understanding between microstructure evolution and mechanical properties is important for designing new high burnup nuclear fuels. In this presentation we discuss the use of picosecond ultrasonics to measure localize changes in mechanical properties of fuel surrogates. We develop measurement techniques that can be applied to investigate heterogeneous elastic properties caused by localize changes in chemistry, grain microstructure caused by recrystallization, and mechanical properties of small samples prepared using focused ion beam sample preparation. Emphasis is placed on understanding the relationship between microstructure and mechanical properties

David Hurley; Marat Khafizov; Farhad Farzbod; Eric Burgett

2013-05-01T23:59:59.000Z

403

Uranium Marketing Annual Report - Release Date: May 31, 2011  

Gasoline and Diesel Fuel Update (EIA)

8. Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors by year, 2008-2012 8. Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors by year, 2008-2012 thousand pounds U3O8 equivalent Origin of Uranium 2008 2009 2010 2011 P2012 Domestic-Origin Uranium 6,228 5,588 4,119 4,134 4,825 Foreign-Origin Uranium 45,040 43,766 40,187 46,809 44,657 Total 51,268 49,354 44,306 50,943 49,483 P = Preliminary data. Final 2011 fuel assembly data reported in the 2012 survey. Notes: Includes only unirradiated uranium in new fuel assemblies loaded into reactors during the year. Does not include uranium removed from reactors that subsequently will be reloaded. Totals may not equal sum of components because of independent rounding. Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual Survey" (2009

404

NUCLEAR ISLANDS International Leasing  

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

ISLANDS ISLANDS International Leasing of Nuclear Fuel Cycle Sites to Provide Enduring Assurance of Peaceful Use Christopher E. Paine and Thomas B. Cochran Current International Atomic Energy Agency safeguards do not provide adequate protection against the diversion to military use of materials or technology from certain types of sensitive nuclear fuel cycle facilities. In view of highly enriched uranium's relatively greater ease of use as a nuclear explosive material than plutonium and the significant diseconomies of commercial spent fuel reprocessing, this article focuses on the need for improved international controls over uranium enrichment facilities as the proximate justification for creation of an International Nuclear Fuel Cycle Association (INFCA). In principle, the proposal is equally applicable to alleviating the proliferation concerns provoked by nuclear fuel

405

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels  

SciTech Connect (OSTI)

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear fuels are critical to understand the burnup, and thus the fuel efficiency.

Lu, Hongbing; Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

2014-01-09T23:59:59.000Z

406

New York Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","5,271",13.4,"41,870",30.6 "Coal","2,781",7.1,"13,583",9.9 "Hydro and Pumped Storage","5,714",14.5,"24,942",18.2 "Natural Gas","17,407",44.2,"48,916",35.7 "Other1",45,0.1,832,0.6 "Other Renewable1","1,719",4.4,"4,815",3.5 "Petroleum","6,421",16.3,"2,005",1.5 "Total","39,357",100.0,"136,962",100.0

407

South Carolina Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","6,486",27.0,"51,988",49.9 "Coal","7,230",30.1,"37,671",36.2 "Hydro and Pumped Storage","4,006",16.7,"1,442",1.4 "Natural Gas","5,308",22.1,"10,927",10.5 "Other1","-","-",61,0.1 "Other Renewable1",284,1.2,"1,873",1.8 "Petroleum",670,2.8,191,0.2 "Total","23,982",100.0,"104,153",100.0

408

Nuclear reactor vessel fuel thermal insulating barrier  

DOE Patents [OSTI]

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19T23:59:59.000Z

409

North Carolina Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,958",17.9,"40,740",31.7 "Coal","12,766",46.1,"71,951",55.9 "Hydro and Pumped Storage","2,042",7.4,"4,757",3.7 "Natural Gas","6,742",24.4,"8,447",6.6 "Other1",50,0.2,407,0.3 "Other Renewable1",543,2.0,"2,083",1.6 "Petroleum",573,2.1,293,0.2 "Total","27,674",100.0,"128,678",100.0

410

EIS-0306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel |  

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

306: Treatment and Management of Sodium-Bonded Spent Nuclear 306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel EIS-0306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel Summary This EIS evaluates the potential environmental impacts of the proposed electrometallurgical treatment of DOE-owned sodium bonded spent nuclear fuel in the Fuel Conditioning Facility at Argonne National Laboratory-West (ANL-W). Public Comment Opportunities None available at this time. Documents Available for Download September 19, 2000 EIS-0306: Record of Decision Treatment and Management of Sodium-Bonded Spent Nuclear Fuel July 1, 2000 EIS-0306: Final Environmental Impact Statement Treatment and Management of Sodium-Bonded Spent Nuclear Fuel July 1, 1999 EIS-0306: Draft Environmental Impact Statement Treatment of Sodium-Bonded Spent Nuclear Fuel

411

U.S. Commits $14 million to U.S. - Ukraine Nuclear Fuel Qualification  

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

Commits $14 million to U.S. - Ukraine Nuclear Fuel Commits $14 million to U.S. - Ukraine Nuclear Fuel Qualification Project U.S. Commits $14 million to U.S. - Ukraine Nuclear Fuel Qualification Project March 15, 2007 - 10:55am Addthis KYIV, Ukraine - U.S. Department of Energy Deputy Secretary Clay Sell today announced that the United States will invest $14 million to provide 42 nuclear fuel assemblies to the South Ukraine Nuclear Power Plant under the U.S.-Ukraine Nuclear Fuel Qualification Project (UNFQP). In an agreement reached last week, Westinghouse Electric Company will manufacture nuclear fuel assemblies, which account for one-fourth of the fuel that powers a reactor for up to four years of operation. Deputy Secretary Sell is in Kyiv today to meet with top Ukrainian officials and U.S. business leaders to promote diversity of energy sources, greater energy efficiency,

412

Categorization of Used Nuclear Fuel Inventory in Support of a Comprehensive  

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

Categorization of Used Nuclear Fuel Inventory in Support of a Categorization of Used Nuclear Fuel Inventory in Support of a Comprehensive National Nuclear Fuel Cycle Strategy Categorization of Used Nuclear Fuel Inventory in Support of a Comprehensive National Nuclear Fuel Cycle Strategy The Office of Nuclear Energy has conducted a technical review and assessment of the total current inventory [~70,150 MTHM as of 2011] of domestic discharged used nuclear fuel (UNF) and estimated an amount to be considered for retention in support of research, development, demonstration, and national security interests. The study recognizes that: 1) access to some amount of UNF is needed to support RD&D; 2) the two principal options for addressing UNF management are geologic disposal and recycling, and 3) U.S. nuclear power plants will continue to discharge

413

Department of Energy Awards $15 Million for Nuclear Fuel Cycle Technology  

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

15 Million for Nuclear Fuel Cycle 15 Million for Nuclear Fuel Cycle Technology Research and Development Department of Energy Awards $15 Million for Nuclear Fuel Cycle Technology Research and Development August 1, 2008 - 2:40pm Addthis WASHINGTON, DC - The U.S. Department of Energy (DOE) today announced it will award up to $15 million to 34 research organizations as part of the Department's Advanced Fuel Cycle Initiative (AFCI). AFCI is the Department's nuclear energy research and development program supporting the long-term goals and objectives of the United States' nuclear energy policy. These projects will provide necessary data and analyses to further U.S. nuclear fuel cycle technology development, meet the need for advanced nuclear energy production and help to close the nuclear fuel cycle

414

EM Safely and Efficiently Manages Spent Nuclear Fuel | Department of Energy  

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

Services » Waste Management » Nuclear Materials & Waste » EM Services » Waste Management » Nuclear Materials & Waste » EM Safely and Efficiently Manages Spent Nuclear Fuel EM Safely and Efficiently Manages Spent Nuclear Fuel Dry storage casks at Idaho National Laboratory can safely house spent nuclear fuel for decades. Dry storage casks at Idaho National Laboratory can safely house spent nuclear fuel for decades. EM's mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal in a geologic repository. Previously, the Office of Environmental Management's (EM) mission had included the safe and efficient management of its spent nuclear fuel (SNF) and preparation for its disposal in a geologic repository. However, in May 2009, the planned geologic repository at Yucca Mountain was cancelled. The

415

Detachable connection for a nuclear reactor fuel assembly  

DOE Patents [OSTI]

A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engageable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA)

1986-01-01T23:59:59.000Z

416

Financing Strategies For A Nuclear Fuel Cycle Facility  

SciTech Connect (OSTI)

To help meet the nation’s energy needs, recycling of partially used nuclear fuel is required to close the nuclear fuel cycle, but implementing this step will require considerable investment. This report evaluates financing scenarios for integrating recycling facilities into the nuclear fuel cycle. A range of options from fully government owned to fully private owned were evaluated using DPL (Decision Programming Language 6.0), which can systematically optimize outcomes based on user-defined criteria (e.g., lowest lifecycle cost, lowest unit cost). This evaluation concludes that the lowest unit costs and lifetime costs are found for a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. However, this does not mean that the facilities should necessarily be constructed and operated by the government. The costs for hybrid combinations of public and private (commercial) financed options can compete under some circumstances with the costs of the government option. This analysis shows that commercial operations have potential to be economical, but there is presently no incentive for private industry involvement. The Nuclear Waste Policy Act (NWPA) currently establishes government ownership of partially used commercial nuclear fuel. In addition, the recently announced Global Nuclear Energy Partnership (GNEP) suggests fuels from several countries will be recycled in the United States as part of an international governmental agreement; this also assumes government ownership. Overwhelmingly, uncertainty in annual facility capacity led to the greatest variations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; and the annual operating costs, forgiveness of debt, and overnight costs dominate the costs computed for the government case. The uncertainty in operations, leading to lower than optimal processing rates (or annual plant throughput), is the most detrimental issue to achieving low unit costs. Conversely, lowering debt interest rates and the required return on investments can reduce costs for private industry.

David Shropshire; Sharon Chandler

2006-07-01T23:59:59.000Z

417

Nuclear fuel reprocessing and the problems of safeguarding against the spread of nuclear weapons  

SciTech Connect (OSTI)

In 1977, the executive branch reversed its long-standing support for nuclear fuel reprocessing, primarily because of the rick of spreading nuclear weapons. GAO reviewed safeguards technology designed to reduce such risks in Federal reprocessing facilities and found that concerns are warranted. Material in sufficient quantities to construct a nuclear weapon could be diverted and go undetected for a long time. Effective international control and safeguards over the production, storage, and use of separated plutonium are lacking. The United States should increase its efforts to: develop and ensure the use of effective safeguards for reprocessing facilities; and establish, in conjunction with major nuclear fuel users, suppliers, and reprocessors, an international system to control the storage and use of excess plutonium.

Staats, E.B.

1980-03-18T23:59:59.000Z

418

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Health and Safety Aspects of Pro- posed Nuclear, Geothermal, and Fossil-FuelHEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUELHealth and Safety Impacts of Nuclear, Geothermal, and Fossil- Fuel

Nero, A.V.

2010-01-01T23:59:59.000Z

419

The breeder reactor: a fossil fuel viewpoint  

Science Journals Connector (OSTI)

... elegant and simple: to generate electricity and, at the same time, to produce additional fuel from the uranium discarded by the existing thermal reactor system. Without the breeder reactor, ... seems likely that the role of nuclear energy will begin to be constrained by the price and availability of uranium at about the turn of the century. There is, however ...

David Merrick

1976-12-16T23:59:59.000Z

420

Final environmental statement related to the Western Nuclear, Inc. , Split Rock Uranium Mill (Fremont County, Wyoming)  

SciTech Connect (OSTI)

The proposed action is the renewal of Source Material License SUA-56 (with amendments) issued to Western Nuclear, Inc. (WNI), for the operation of the Split Rock Uranium Mill near Jeffrey City and the Green Mountain Ion-Exchange Facility, both in Fremont County, Wyoming. The license also permits possession of material from past operations at four ancillary facilities in the Gas Hills mining area - the Bullrush, Day-Loma, Frazier-Lamac, and Rox sites (Docket No. 40-1162). However, although heap leaching operations were previously authorized at Frazier-Lamac, there has never been any processing of material at this site. The Split Rock mill is an acid-leach, ion-exchange and solvent-extraction uranium-ore processing mill with a design capacity of 1540 MT (1700 tons) of ore per day. WNI has proposed by license amendment request to increase the storage capacity of the tailings ponds in order to permit the continuation of present production rates of U/sub 3/O/sub 8/ through 1996 using lower-grade ores.

Not Available

1980-02-01T23:59:59.000Z

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421

Integrated Data Base report--1993: U.S. spent nuclear fuel and radioactive waste inventories, projections, and characteristics. Revision 10  

SciTech Connect (OSTI)

The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and DOE spent nuclear fuel; also, commercial and US government-owned radioactive wastes through December 31, 1993. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 256 refs., 38 figs., 141 tabs.

Not Available

1994-12-01T23:59:59.000Z

422

Pyroprocessing oxide spent nuclear fuels for efficient disposal  

SciTech Connect (OSTI)

Pyrochemical processing as a means for conditioning spent nuclear fuels for disposal offers significant advantages over the direct disposal option. The advantages include reduction in high-level waste volume; conversion of most of the high-level waste to a low-level waste in which nearly all the transuranics (TRU) have been removed; and incorporation of the TRUs into a stable, highly radioactive waste form suitable for interim storage, ultimate destruction, or repository disposal. The lithium process has been under development at Argonne National Laboratory for use in pyrochemical conditioning of spent fuel for disposal. All of the process steps have been demonstrated in small-scale (0.5-kg simulated spent fuel) experiments. Engineering-scale (20-kg simulated spent fuel) demonstration of the process is underway, and small-scale experiments have been conducted with actual spent fuel from a light water reactor (LWR). The lithium process is simple, operates at relatively low temperatures, and can achieve high decontamination factors for the TRU elements. Ordinary materials, such as carbon steel, can be used for process containment.

McPheeters, C.C.; Pierce, R.D.; Mulcahey, T.P. [Argonne National Lab., IL (United States). Chemical Technology Div.

1994-12-31T23:59:59.000Z

423

New Prototype Safeguards Technology Offers Improved Confidence and Automation for Uranium Enrichment Facilities  

SciTech Connect (OSTI)

An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.

Brim, Cornelia P.

2013-03-04T23:59:59.000Z

424

Synergistic energy conversion processes using nuclear energy and fossil fuels  

Science Journals Connector (OSTI)

This paper reviews the methods of producing energy carriers, such as electricity, hydrocarbons and hydrogen, by utilising both nuclear energy and fossil fuels synergistically. There are many possibilities for new, innovative, synergistic processes, which combine chemical and nuclear systems for efficient, clean and economical production of energy carriers. Besides the individual processes by each form of energy to produce the energy carriers, the synergistic processes which use two primary energies to produce the energy carriers will become important with the features of resource saving, CO2 emission reduction and economic production, due to the higher conversion efficiency and low cost of nuclear heat. The synergistic processes will be indispensable to the 21st century, when efficient best-mixed supplies of available primary energies are crucial.

Masao Hori

2009-01-01T23:59:59.000Z

425

Foreign Research Reactor Spent Nuclear Fuel Acceptance Program  

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

Global Threat Reduction Initiative: Global Threat Reduction Initiative: U.S. Nuclear Remove Program Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance 2007 DOE TEC Meeting Chuck Messick DOE/NNSA/SRS 2 Contents * Program Objective and Policy * Program implementation status * Shipment Information * Operational Logistics * Lessons Learned * Conclusion 3 U.S. Nuclear Remove Program Objective * To play a key role in the Global Threat Reduction Remove Program supporting permanent threat reduction by accepting program eligible material. * Works in conjunction with the Global Threat Reduction Convert Program to accept program eligible material as an incentive to core conversion providing a disposition path for HEU and LEU during the life of the Acceptance Program. 4 Reasons for the Policy

426

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel  

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

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the