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Sample records for nuclear driver fuel

  1. Nuclear Fuels.

    ERIC Educational Resources Information Center

    Nash, J. Thomas

    1983-01-01

    Trends in and factors related to the nuclear industry and nuclear fuel production are discussed. Topics addressed include nuclear reactors, survival of the U.S. uranium industry, production costs, budget cuts by the Department of Energy and U.S. Geological survey for resource studies, mining, and research/development activities. (JN)

  2. Nuclear reactor composite fuel assembly

    DOEpatents

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  3. Fuel-Cell Drivers Wanted

    ERIC Educational Resources Information Center

    Clark, Todd; Jones, Rick

    2004-01-01

    While the political climate seems favorable for the development of fuel-cell vehicles for personal transportation, the market's demand may not be so favorable. Nonetheless, middle level students will be the next generation of drivers and voters, and they need to be able to make informed decisions regarding the nation's energy and transportation…

  4. Vented nuclear fuel element

    DOEpatents

    Grossman, Leonard N.; Kaznoff, Alexis I.

    1979-01-01

    A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

  5. Nuclear Fuel Reprocessing

    SciTech Connect

    Michael F. Simpson; Jack D. Law

    2010-02-01

    This is an a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. No formal abstract was required for the article. The full article will be attached.

  6. Nuclear fuel element

    DOEpatents

    Zocher, Roy W.

    1991-01-01

    A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

  7. Nuclear fuel element

    DOEpatents

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  8. Fresh nuclear fuel measurements at Ukrainian nuclear power plants

    SciTech Connect

    Kuzminski, Jozef; Ewing, Tom; Dickman, Debbie; Gavrilyuk, Victor; Drapey, Sergey; Kirischuk, Vladimir; Strilchuk, Nikolay

    2009-01-01

    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.

  9. Nuclear fuel pin scanner

    DOEpatents

    Bramblett, Richard L.; Preskitt, Charles A.

    1987-03-03

    Systems and methods for inspection of nuclear fuel pins to determine fiss loading and uniformity. The system includes infeed mechanisms which stockpile, identify and install nuclear fuel pins into an irradiator. The irradiator provides extended activation times using an approximately cylindrical arrangement of numerous fuel pins. The fuel pins can be arranged in a magazine which is rotated about a longitudinal axis of rotation. A source of activating radiation is positioned equidistant from the fuel pins along the longitudinal axis of rotation. The source of activating radiation is preferably oscillated along the axis to uniformly activate the fuel pins. A detector is provided downstream of the irradiator. The detector uses a plurality of detector elements arranged in an axial array. Each detector element inspects a segment of the fuel pin. The activated fuel pin being inspected in the detector is oscillated repeatedly over a distance equal to the spacing between adjacent detector elements, thereby multiplying the effective time available for detecting radiation emissions from the activated fuel pin.

  10. Processing of driver fuel assemblies at FFTF

    SciTech Connect

    Danko, A.D.; Hicks, D.F.; Arneson, S.O.

    1982-07-01

    The ability to disassemble an irradiated Fast Flux Test Facility (FFTF) Driver Fuel Assembly (DFA) is important both to the continued operation of the FFTF and the future of the Breeder Reactor Program. At the FFTF, DFA's with up to three (3)* kilowatts of decay heat will be placed in the Interim Examination and Maintenance (IEM) Cell for disassembly and nondestructive examination. This process includes sodium removal, duct measurement, duct cutting and pulling, fuel pin removal, and component disposition to other laboratories for destructive examination.

  11. Nuclear Fuel Cycle

    SciTech Connect

    Dale, Deborah J.

    2014-10-28

    These slides will be presented at the training course “International Training Course on Implementing State Systems of Accounting for and Control (SSAC) of Nuclear Material for States with Small Quantity Protocols (SQP),” on November 3-7, 2014 in Santa Fe, New Mexico. The slides provide a basic overview of the Nuclear Fuel Cycle. This is a joint training course provided by NNSA and IAEA.

  12. NUCLEAR FUEL MATERIAL

    DOEpatents

    Goeddel, W.V.

    1962-06-26

    An improved method is given for making the carbides of nuclear fuel material. The metal of the fuel material, which may be a fissile and/or fertile material, is transformed into a silicide, after which the silicide is comminuted to the desired particle size. This silicide is then carburized at an elevated temperature, either above or below the melting point of the silicide, to produce an intimate mixture of the carbide of the fuel material and the carbide of silicon. This mixture of the fuel material carbide and the silicon carbide is relatively stable in the presence of moisture and does not exhibit the highly reactive surface condition which is observed with fuel material carbides made by most other known methods. (AEC)

  13. Nuclear Fuel Reprocessing

    SciTech Connect

    Harold F. McFarlane; Terry Todd

    2013-11-01

    Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore. Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor

  14. Nuclear fuel cycle information workshop

    SciTech Connect

    Not Available

    1983-01-01

    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.

  15. NUCLEAR FUEL COMPOSITION

    DOEpatents

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

    1960-05-31

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

  16. Nuclear fuel element

    DOEpatents

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1980-04-29

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

  17. Nuclear fuel element

    DOEpatents

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

  18. Nuclear reactor fuel element

    DOEpatents

    Johnson, Carl E.; Crouthamel, Carl E.

    1980-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

  19. Nuclear fuel electrorefiner

    DOEpatents

    Ahluwalia, Rajesh K.; Hua, Thanh Q.

    2004-02-10

    The present invention relates to a nuclear fuel electrorefiner having a vessel containing a molten electrolyte pool floating on top of a cadmium pool. An anodic fuel dissolution basket and a high-efficiency cathode are suspended in the molten electrolyte pool. A shroud surrounds the fuel dissolution basket and the shroud is positioned so as to separate the electrolyte pool into an isolated electrolyte pool within the shroud and a bulk electrolyte pool outside the shroud. In operation, unwanted noble-metal fission products migrate downward into the cadmium pool and form precipitates where they are removed by a filter and separator assembly. Uranium values are transported by the cadmium pool from the isolated electrolyte pool to the bulk electrolyte pool, and then pass to the high-efficiency cathode where they are electrolytically deposited thereto.

  20. NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1962-08-14

    A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

  1. Swelling-resistant nuclear fuel

    DOEpatents

    Arsenlis, Athanasios [Hayward, CA; Satcher, Jr., Joe; Kucheyev, Sergei O [Oakland, CA

    2011-12-27

    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.

  2. Risks of nuclear fuel reprocessing

    SciTech Connect

    Durant, W.S.

    1990-01-01

    The Savannah River Site's primary function is the production of weapons materials. It consists of four reactors, two fuel reprocessing facilities, a fuel fabrication facility, a nuclear fuel facility for the Navy and a heavy water recycle facility. Under construction is a facility to convert the site's liquid wastes into borosilicate glass. The topic of this paper is risks of nuclear fuel reprocessing. Also discussed are facility operations. 18 figs.

  3. Advanced Nuclear Fuel Cycle Options

    SciTech Connect

    Roald Wigeland; Temitope Taiwo; Michael Todosow; William Halsey; Jess Gehin

    2010-06-01

    A systematic evaluation has been conducted of the potential for advanced nuclear fuel cycle strategies and options to address the issues ascribed to the use of nuclear power. Issues included nuclear waste management, proliferation risk, safety, security, economics and affordability, and sustainability. The two basic strategies, once-through and recycle, and the range of possibilities within each strategy, are considered for all aspects of the fuel cycle including options for nuclear material irradiation, separations if needed, and disposal. Options range from incremental changes to today’s implementation to revolutionary concepts that would require the development of advanced nuclear technologies.

  4. Nuclear Fuel Cycle & Vulnerabilities

    SciTech Connect

    Boyer, Brian D.

    2012-06-18

    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.

  5. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  6. Spent Nuclear Fuel project, project management plan

    SciTech Connect

    Fuquay, B.J.

    1995-10-25

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  7. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-16

    A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

  8. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

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

    1959-09-15

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

  9. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Anderson, W.F.; Tellefson, D.R.; Shimazaki, T.T.

    1962-04-10

    A plate type fuel element which is particularly useful for organic cooled reactors is described. Generally, the fuel element comprises a plurality of fissionable fuel bearing plates held in spaced relationship by a frame in which the plates are slidably mounted in grooves. Clearance is provided in the grooves to allow the plates to expand laterally. The plates may be rigidly interconnected but are floatingly supported at their ends within the frame to allow for longi-tudinal expansion. Thus, this fuel element is able to withstand large temperature differentials without great structural stresses. (AEC)

  10. Nuclear fuels - Present and future

    NASA Astrophysics Data System (ADS)

    Olander, D.

    2009-06-01

    The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

  11. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1963-06-11

    A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

  12. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-01

    A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

  13. Nuclear Fuels: Promise and Limitations

    SciTech Connect

    Harold F. McFarlane

    2012-03-01

    From 1950 through 1980, scientists, engineers and national leaders confidently predicted an early twenty-first century where fast breeder reactors and commercial nuclear fuel reprocessing were commonplace. Such a scenario seemed necessary for a world with the more than 1000 GWe of nuclear energy needed to meet such an ever-increasing thirst for energy. Thirty years later uranium reserves are increasing on pace with consumption, the growth of nuclear power has been slowed, commercial breeder reactors have yet to enter the marketplace, and less than a handful of commercial reprocessing plants operate. As Nobel Laureate Niels Bohr famously said, “Prediction is very difficult, especially if it’s about the future.” The programme for IChemE’s 2012 conference on the nuclear fuel cycle features a graphic of an idealized nuclear fuel cycle that symbolizes the quest for a closed nuclear fuel cycle featuring careful husbanding of precious resources while minimizing the waste footprint. Progress toward achieving this ideal has been disrupted by technology innovations in the mining and petrochemical industries, as well as within the nuclear industry.

  14. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  15. Analyzing Vehicle Fuel Saving Opportunities through Intelligent Driver Feedback

    SciTech Connect

    Gonder, J.; Earleywine, M.; Sparks, W.

    2012-06-01

    Driving style changes, e.g., improving driver efficiency and motivating driver behavior changes, could deliver significant petroleum savings. This project examines eliminating stop-and-go driving and unnecessary idling, and also adjusting acceleration rates and cruising speeds to ideal levels to quantify fuel savings. Such extreme adjustments can result in dramatic fuel savings of over 30%, but would in reality only be achievable through automated control of vehicles and traffic flow. In real-world driving, efficient driving behaviors could reduce fuel use by 20% on aggressively driven cycles and by 5-10% on more moderately driven trips. A literature survey was conducted of driver behavior influences, and pertinent factors from on-road experiments with different driving styles were observed. This effort highlighted important driver influences such as surrounding vehicle behavior, anxiety over trying to get somewhere quickly, and the power/torque available from the vehicle. Existing feedback approaches often deliver efficiency information and instruction. Three recommendations for maximizing fuel savings from potential drive cycle improvement are: (1) leveraging applications with enhanced incentives, (2) using an approach that is easy and widely deployable to motivate drivers, and (3) utilizing connected vehicle and automation technologies to achieve large and widespread efficiency improvements.

  16. Proliferation Resistant Nuclear Reactor Fuel

    SciTech Connect

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and

  17. Spent nuclear fuel reprocessing modeling

    SciTech Connect

    Tretyakova, S.; Shmidt, O.; Podymova, T.; Shadrin, A.; Tkachenko, V.; Makeyeva, I.; Tkachenko, V.; Verbitskaya, O.; Schultz, O.; Peshkichev, I.

    2013-07-01

    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)

  18. Protected Nuclear Fuel Element

    DOEpatents

    Kittel, J. H.; Schumar, J. F.

    1962-12-01

    A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

  19. Fire resistant nuclear fuel cask

    DOEpatents

    Heckman, Richard C.; Moss, Marvin

    1979-01-01

    The disclosure is directed to a fire resistant nuclear fuel cask employing reversibly thermally expansible bands between adjacent cooling fins such that normal outward flow of heat is not interfered with, but abnormal inward flow of heat is impeded or blocked.

  20. World nuclear fuel cycle requirements 1991

    SciTech Connect

    Not Available

    1991-10-10

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs.

  1. Final Report on the Fuel Saving Effectiveness of Various Driver Feedback Approaches

    SciTech Connect

    Gonder, J.; Earleywine, M.; Sparks, W.

    2011-03-01

    This final report quantifies the fuel-savings opportunities from specific driving behavior changes, identifies factors that influence drivers' receptiveness to adopting fuel-saving behaviors, and assesses various driver feedback approaches.

  2. Fully ceramic nuclear fuel and related methods

    DOEpatents

    Venneri, Francesco; Katoh, Yutai; Snead, Lance Lewis

    2016-03-29

    Various embodiments of a nuclear fuel for use in various types of nuclear reactors and/or waste disposal systems are disclosed. One exemplary embodiment of a nuclear fuel may include a fuel element having a plurality of tristructural-isotropic fuel particles embedded in a silicon carbide matrix. An exemplary method of manufacturing a nuclear fuel is also disclosed. The method may include providing a plurality of tristructural-isotropic fuel particles, mixing the plurality of tristructural-isotropic fuel particles with silicon carbide powder to form a precursor mixture, and compacting the precursor mixture at a predetermined pressure and temperature.

  3. Analysis of IFR driver fuel hot channel factors

    SciTech Connect

    Ku, J.Y.; Chang, L.K.; Mohr, D.

    1994-03-01

    Thermal-hydraulic uncertainty factors for Integral Fast Reactor (IFR) driver fuels have been determined based primarily on the database obtained from the predecessor fuels used in the IFR prototype, Experimental Breeder Reactor II. The uncertainty factors were applied to the channel factors (HCFs) analyses to obtain separate overall HCFs for fuel and cladding for steady-state analyses. A ``semistatistical horizontal method`` was used in the HCFs analyses. The uncertainty factor of the fuel thermal conductivity dominates the effects considered in the HCFs analysis; the uncertainty in fuel thermal conductivity will be reduced as more data are obtained to expand the currently limited database for the IFR ternary metal fuel (U-20Pu-10Zr). A set of uncertainty factors to be used for transient analyses has also been derived.

  4. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    DOEpatents

    Stengel, F.G.

    1963-12-24

    A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

  5. Long-term global nuclear energy and fuel cycle strategies

    SciTech Connect

    Krakowski, R.A.

    1997-09-24

    The Global Nuclear Vision Project is examining, using scenario building techniques, a range of long-term nuclear energy futures. The exploration and assessment of optimal nuclear fuel-cycle and material strategies is an essential element of the study. To this end, an established global E{sup 3} (energy/economics/environmental) model has been adopted and modified with a simplified, but comprehensive and multi-regional, nuclear energy module. Consistent nuclear energy scenarios are constructed using this multi-regional E{sup 3} model, wherein future demands for nuclear power are projected in price competition with other energy sources under a wide range of long-term demographic (population, workforce size and productivity), economic (price-, population-, and income-determined demand for energy services, price- and population-modified GNP, resource depletion, world-market fossil energy prices), policy (taxes, tariffs, sanctions), and top-level technological (energy intensity and end-use efficiency improvements) drivers. Using the framework provided by the global E{sup 3} model, the impacts of both external and internal drivers are investigated. The ability to connect external and internal drivers through this modeling framework allows the study of impacts and tradeoffs between fossil- versus nuclear-fuel burning, that includes interactions between cost, environmental, proliferation, resource, and policy issues.

  6. Compositions and methods for treating nuclear fuel

    DOEpatents

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

    2013-08-13

    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.

  7. Compositions and methods for treating nuclear fuel

    DOEpatents

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

    2014-01-28

    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.

  8. Nuclear fuel microsphere gamma analyzer

    DOEpatents

    Valentine, Kenneth H.; Long, Jr., Ernest L.; Willey, Melvin G.

    1977-01-01

    A gamma analyzer system is provided for the analysis of nuclear fuel microspheres and other radioactive particles. The system consists of an analysis turntable with means for loading, in sequence, a plurality of stations within the turntable; a gamma ray detector for determining the spectrum of a sample in one section; means for analyzing the spectrum; and a receiver turntable to collect the analyzed material in stations according to the spectrum analysis. Accordingly, particles may be sorted according to their quality; e.g., fuel particles with fractured coatings may be separated from those that are not fractured, or according to other properties.

  9. HIGH DENSITY NUCLEAR FUEL COMPOSITION

    DOEpatents

    Litton, F.B.

    1962-07-17

    ABS>A nuclear fuel consisting essentially of uranium monocarbide and containing 2.2 to 4.6 wt% carbon, 0.1 to 2.3 wt% oxygen, 0.05 to 2.5 wt% nitrogen, and the balance uranium was developed. The maximum oxygen content was less than one-half the carbon content by weight and the carbon, oxygen, and nitrogen are present as a single phase substituted solid solution of UC, C, O, and N. A method of preparing the fuel composition is described. (AEC)

  10. Nuclear fuel pellet transfer escalator

    SciTech Connect

    Huggins, T.B. Sr.; Roberts, E.; Edmunds, M.O.

    1991-09-17

    This patent describes a nuclear fuel pellet escalator for loading nuclear fuel pellets into a sintering boat. It comprises a generally horizontally-disposed pellet transfer conveyor for moving pellets in single file fashion from a receiving end to a discharge end thereof, the conveyor being mounted about an axis at its receiving end for pivotal movement to generally vertically move its discharge end toward and away from a sintering boat when placed below the discharge end of the conveyor, the conveyor including an elongated arm swingable vertically about the axis and having an elongated channel recessed below an upper side of the arm and extending between the receiving and discharge ends of the conveyor; a pellet dispensing chute mounted to the arm of the conveyor at the discharge end thereof and extending therebelow such that the chute is carried at the discharge end of the conveyor for generally vertical movement therewith toward and away from the sintering boat.

  11. The Drivers of Indias Nuclear Weapons Program

    DTIC Science & Technology

    2014-06-01

    135 Ankit Panda , “India Inches Closer to Credible Nuclear Triad with K-4 SLBM Test,” The Diplomat, May 13, 2014, http://thediplomat.com/2014/05/india...212 Panda , “India Inches Closer to Credible Nuclear Triad with K-4 SLBM Test.” 213 Rajagopalan, “The Logic of Assured Retaliation...Ankit Panda , “The Nuclear Problem in India-Japan Relations,” The Diplomat, October 31, 2013, http://thediplomat.com/2013/10/the-nuclear-problem-in-india

  12. FUEL COMPOSITION FOR NUCLEAR REACTORS

    DOEpatents

    Andersen, J.C.

    1963-08-01

    A process for making refractory nuclear fuel elements involves heating uranium and silicon powders in an inert atmosphere to 1600 to 1800 deg C to form USi/sub 3/; adding silicon carbide, carbon, 15% by weight of nickel and aluminum, and possibly also molybdenum and silicon powders; shaping the mixture; and heating to 1700 to 2050 deg C again in an inert atmosphere. Information on obtaining specific compositions is included. (AEC)

  13. Method for producing nuclear fuel

    DOEpatents

    Haas, Paul A.

    1983-01-01

    Nuclear fuel is made by contacting an aqueous solution containing an actinide salt with an aqueous solution containing ammonium hydroxide, ammonium oxalate, or oxalic acid in an amount that will react with a fraction of the actinide salt to form a precipitate consisting of the hydroxide or oxalate of the actinide. A slurry consisting of the precipitate and solution containing the unreacted actinide salt is formed into drops which are gelled, calcined, and pressed to form pellets.

  14. Method for producing nuclear fuel

    SciTech Connect

    Haas, P.A.

    1981-04-24

    Nuclear fuel is made by contacting an aqueous solution containing an actinide salt with an aqueous solution containing ammonium hydroxide, ammonium oxalate, or oxalic acid in an amount that will react with a fraction of the actinide salt to form a precipitate consisting of the hydroxide or oxalate of the actinide. A slurry consisting of the precipitate and solution containing the unreacted actinide salt is formed into drops which are gelled, calcined, and pressed to form pellets.

  15. Nuclear fuel assembly wear sleeve

    SciTech Connect

    Cadwell, D.J.; Kmonk, S.

    1983-03-08

    An improved control rod guide tube for use in a fuel assembly in a nuclear reactor. The guide tube extends the complete length of the fuel assembly and has its upper end fastened in a cylindrical housing by swaging the guide tube material into grooves formed in the housing walls. To eliminate wear on the guide tube inner walls caused by hydraulic induced vibratory forces on a control rod adapted to move therein, a thin-walled chrome plated sleeve is threaded into the top end of the guide thimble and extends downwardly a distance sufficient to be engaged by the control rod during reactor operation. The sleeve serves as a highly resistant wear surface between the control rod and walls on the guide tube in the fuel assembly.

  16. Investigation of vehicle and driver aggressivity and relation to fuel economy testing

    NASA Astrophysics Data System (ADS)

    Stichter, Jonathan Seth

    As vehicle technologies continue to improve it is becoming more evident one of the last major factors impacting fuel economy left today is the driver. In this study the driver is defined as the operator of a vehicle and the difference between driving styles of the driver and vehicle is defined as aggressivity. Driver aggressivity is proven to have a substantial impact on fuel economy in many studies. Many fuel economy tests have been created, all to measure the fuel efficiency of today's vehicles and their related technologies. These tests typically require that the drivers be trained or experienced in fuel economy testing unless the impact of the driver on fuel economy is the variable being tested. It is also recommended, for certain tests, that the driver stay with the same vehicle for the tests entirety. Although these are the requirements, having the same trained drivers for the entirety of a fuel economy test may not always be a viable option. This leads to the question of, what impact can a set of drivers, who are asked to drive the same, have on fuel consumption during a fuel consumption test? The SAE J1321 Type II Fuel Consumption Test Procedure was followed on two identical trucks with two drivers that were untrained in fuel economy testing in order to answer this question. It was found in this particular study that the driver variability can impose up to a 10% fuel economy difference on shorter distance routes where the driver is kept the same. By increasing the distance of the route and swapping drivers variability in fuel economy reduced to 5%. It was shown by this particular test that the impact of the driver when asked to drive the same is minimal compared to real world results of up to 30%. A larger data set and more testing is still necessary to completely understand and validate the impact of the driver on fuel economy testing.

  17. Welding fixture for nuclear fuel pin cladding assemblies

    DOEpatents

    Oakley, D.J.; Feld, S.H.

    1984-02-22

    A welding fixture is described for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

  18. Welding fixture for nuclear fuel pin cladding assemblies

    DOEpatents

    Oakley, David J.; Feld, Sam H.

    1986-01-01

    A welding fixture for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

  19. Nuclear core and fuel assemblies

    DOEpatents

    Downs, Robert E.

    1981-01-01

    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.

  20. Variants of closing the nuclear fuel cycle

    SciTech Connect

    Andrianova, E. A. Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V.

    2015-12-15

    Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

  1. Survey of Dynamic Simulation Programs for Nuclear Fuel Reprocessing

    SciTech Connect

    Troy J. Tranter; Daryl R. Haefner

    2008-06-01

    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.

  2. The IFR modern nuclear fuel cycle

    SciTech Connect

    Hannum, W.H.

    1991-01-01

    Nuclear power is an essential component of the world's energy supply. The IFR program, by returning to fundamentals, offers a fresh approach to closing the nuclear fuel cycle. This closed fuel cycle represents the ultimate in efficient resource utilization and environmental accountability. 35 refs., 2 tabs.

  3. Nuclear fuel: a new market dynamic

    SciTech Connect

    Kee, Edward D.

    2007-12-15

    After almost 20 years of low nuclear fuel prices, buyers have come to expect that these low and stable nuclear fuel prices will continue. This conventional wisdom may not reflect the significant changes and higher prices that growing demand, and the end of secondary sources of uranium and enrichment, will bring. (author)

  4. Monitoring arrangement for vented nuclear fuel elements

    DOEpatents

    Campana, Robert J.

    1981-01-01

    In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

  5. Used Nuclear Fuel: From Liability to Benefit

    NASA Astrophysics Data System (ADS)

    Orbach, Raymond L.

    2011-03-01

    Nuclear power has proven safe and reliable, with operating efficiencies in the U.S. exceeding 90%. It provides a carbon-free source of electricity (with about a 10% penalty arising from CO2 released from construction and the fuel cycle). However, used fuel from nuclear reactors is highly toxic and presents a challenge for permanent disposal -- both from technical and policy perspectives. The half-life of the ``bad actors'' is relatively short (of the order of decades) while the very long lived isotopes are relatively benign. At present, spent fuel is stored on-site in cooling ponds. Once the used fuel pools are full, the fuel is moved to dry cask storage on-site. Though the local storage is capable of handling used fuel safely and securely for many decades, the law requires DOE to assume responsibility for the used fuel and remove it from reactor sites. The nuclear industry pays a tithe to support sequestration of used fuel (but not research). However, there is currently no national policy in place to deal with the permanent disposal of nuclear fuel. This administration is opposed to underground storage at Yucca Mountain. There is no national policy for interim storage---removal of spent fuel from reactor sites and storage at a central location. And there is no national policy for liberating the energy contained in used fuel through recycling (separating out the fissionable components for subsequent use as nuclear fuel). A ``Blue Ribbon Commission'' has been formed to consider alternatives, but will not report until 2012. This paper will examine alternatives for used fuel disposition, their drawbacks (e.g. proliferation issues arising from recycling), and their benefits. For recycle options to emerge as a viable technology, research is required to develop cost effective methods for treating used nuclear fuel, with attention to policy as well as technical issues.

  6. World nuclear fuel cycle requirements 1990

    SciTech Connect

    Not Available

    1990-10-26

    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.

  7. Fuel handling apparatus for a nuclear reactor

    DOEpatents

    Hawke, Basil C.

    1987-01-01

    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.

  8. Integrated process for reprocessing spent nuclear fuel

    SciTech Connect

    Forsberg, C.W.

    1991-03-06

    This invention is comprised of a process for recovering nuclear fuel from spent fuel assemblies that employs a single canister process container. The cladding and fuel are oxidized in the container, the fuel is dissolved and removed from the container for separation from the aqueous phase, the aqueous phase containing radioactive waste is returned to the container. This container is also the disposal vessel. Add solidification agents and compress container for long term storage.

  9. Evolving an acceptable nuclear power fuel cycle

    SciTech Connect

    Steinberg, M.

    1986-10-01

    The following issues are examined: long-term safe nuclear power plant operation; acceptable nuclear waste management and, mainly, high-level waste management; and provision for long-term fissile fuel supply in a long-term nuclear fission economy. (LM)

  10. Nuclear fuel elements having a composite cladding

    DOEpatents

    Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

    1983-09-20

    An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

  11. CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

    DOEpatents

    Weems, S.J.

    1963-09-24

    >A rib arrangement and an end construction for nuclearfuel elements laid end to end in a coolant tube are described. The rib arrangement is such that each fuel element, when separated from other fuel elements, fits loosely in the coolant tube and so can easily be inserted or withdrawn from the tube. The end construction of the fuel elements is such that the fuel elements when assembled end to end are keyed against relative rotation, and the ribs of each fuel element cooperate with the ribs of the adjacent fuel elements to give the assembled fuel elements a tight fit with the coolant tube. (AEC)

  12. International Nuclear Fuel Cycle Fact Book

    SciTech Connect

    Leigh, I.W.; Patridge, M.D.

    1991-05-01

    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 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 OECN/NEA activities reports; not reflect any one single source but frequently represent a consolidation/combination of information.

  13. Annotated Bibliography for Drying Nuclear Fuel

    SciTech Connect

    Rebecca E. Smith

    2011-09-01

    Internationally, the nuclear industry is represented by both commercial utilities and research institutions. Over the past two decades many of these entities have had to relocate inventories of spent nuclear fuel from underwater storage to dry storage. These efforts were primarily prompted by two factors: insufficient storage capacity (potentially precipitated by an open-ended nuclear fuel cycle) or deteriorating quality of existing underwater facilities. The intent of developing this bibliography is to assess what issues associated with fuel drying have been identified, to consider where concerns have been satisfactorily addressed, and to recommend where additional research would offer the most value to the commercial industry and the U. S. Department of Energy.

  14. Method for shearing spent nuclear fuel assemblies

    DOEpatents

    Weil, Bradley S.; Watson, Clyde D.

    1977-01-01

    A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

  15. Overview of the nuclear fuel cycle

    SciTech Connect

    Leuze, R.E.

    1982-01-01

    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.

  16. Overview of the nuclear fuel cycle

    SciTech Connect

    Leuze, R.E.

    1981-01-01

    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.

  17. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-11-21

    A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

  18. Apparatus for shearing spent nuclear fuel assemblies

    DOEpatents

    Weil, Bradley S.; Metz, III, Curtis F.

    1980-01-01

    A method and apparatus are described for shearing spent nuclear fuel assemblies of the type comprising an array of fuel pins disposed within an outer metal shell or shroud. A spent fuel assembly is first compacted in a known manner and then incrementally sheared using fixed and movable shear blades having matched laterally projecting teeth which slidably intermesh to provide the desired shearing action. Incremental advancement of the fuel assembly after each shear cycle is limited to a distance corresponding to the lateral projection of the teeth to ensure fuel assembly breakup into small uniform segments which are amenable to remote chemical processing.

  19. Composite construction for nuclear fuel containers

    DOEpatents

    Cheng, Bo-Ching [Fremont, CA; Rosenbaum, Herman S [Fremont, CA; Armijo, Joseph S [Saratoga, CA

    1987-01-01

    An improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof.

  20. Composite construction for nuclear fuel containers

    DOEpatents

    Cheng, B. C.; Rosenbaum, H. S.; Armijo, J. S.

    1987-04-21

    Disclosed is an improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof. 1 fig.

  1. Fuel assembly for nuclear reactors

    DOEpatents

    Creagan, Robert J.; Frisch, Erling

    1977-01-01

    A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

  2. Spent Nuclear Fuel Project Technical Databook

    SciTech Connect

    Reilly, M.A.

    1998-10-23

    The Spent Nuclear Fuel (SNF) Project Technical Databook is developed for use as a common authoritative source of fuel behavior and material parameters in support of the Hanford SNF Project. The Technical Databook will be revised as necessary to add parameters as their Databook submittals become available.

  3. Pyrolytic carbon-coated nuclear fuel

    DOEpatents

    Lindemer, Terrence B.; Long, Jr., Ernest L.; Beatty, Ronald L.

    1978-01-01

    An improved nuclear fuel kernel having at least one pyrolytic carbon coating and a silicon carbon layer is provided in which extensive interaction of fission product lanthanides with the silicon carbon layer is avoided by providing sufficient UO.sub.2 to maintain the lanthanides as oxides during in-reactor use of said fuel.

  4. 77 FR 19278 - Informational Meeting on Nuclear Fuel Cycle Options

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-30

    ... Informational Meeting on Nuclear Fuel Cycle Options AGENCY: Office of Fuel Cycle Technologies, Office of Nuclear.... At this meeting, input is being sought from participants knowledgeable in nuclear fuel cycles... evaluation and screening of nuclear fuel cycle options in 2013. At this meeting, input is being sought...

  5. Rack for storing spent nuclear fuel elements

    DOEpatents

    Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

    1978-06-20

    A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

  6. Dry Processing of Used Nuclear Fuel

    SciTech Connect

    K. M. Goff; M. F. Simpson

    2009-09-01

    Dry (non-aqueous) separations technologies have been used for treatment of used nuclear fuel since the 1960s, and they are still being developed and demonstrated in many countries. Dry technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. Within the Department of Energy’s Advanced Fuel Cycle Initiative, an electrochemical process employing molten salts is being developed for recycle of fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. Much of the development of this technology is based on treatment of used Experimental Breeder Reactor II (EBR-II) fuel, which is metallic. Electrochemical treatment of the EBR-II fuel has been ongoing in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory since 1996. More than 3.8 metric tons of heavy metal of metallic fast reactor fuel have been treated using this technology. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including high-level waste work. A historic perspective on the background of dry processing will also be provided.

  7. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Dickson, J.J.

    1963-09-24

    A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

  8. Spent Nuclear Fuel Transport Reliability Study

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Jiang, Hao

    2016-01-01

    This conference paper was orignated and shorten from the following publisehd PTS documents: 1. Jy-An Wang, Hao Jiang, and Hong Wang, Dynamic Deformation Simulation of Spent Nuclear Fuel Assembly and CIRFT Deformation Sensor Stability Investigation, ORNL/SPR-2015/662, November 2015. 2. Jy-An Wang, Hong Wang, Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications, NUREG/CR-7198, ORNL/TM-2014/214, May 2015. 3. Jy-An Wang, Hong Wang, Hao Jiang, Yong Yan, Bruce Bevard, Spent Nuclear Fuel Vibration Integrity Study 16332, WM2016 Conference, March 6 10, 2016, Phoenix, Arizona.

  9. Nuclear Fuels & Materials Spotlight Volume 4

    SciTech Connect

    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

    2014-04-01

    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.

  10. Nuclear fuel in a reactor accident.

    PubMed

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

    2012-03-09

    Nuclear accidents that lead to melting of a reactor core create heterogeneous materials containing hundreds of radionuclides, many with short half-lives. The long-lived fission products and transuranium elements within damaged fuel remain a concern for millennia. Currently, accurate fundamental models for the prediction of release rates of radionuclides from fuel, especially in contact with water, after an accident remain limited. Relatively little is known about fuel corrosion and radionuclide release under the extreme chemical, radiation, and thermal conditions during and subsequent to a nuclear accident. We review the current understanding of nuclear fuel interactions with the environment, including studies over the relatively narrow range of geochemical, hydrological, and radiation environments relevant to geological repository performance, and discuss priorities for research needed to develop future predictive models.

  11. Dry Transfer Systems for Used Nuclear Fuel

    SciTech Connect

    Brett W. Carlsen; Michaele BradyRaap

    2012-05-01

    The potential need for a dry transfer system (DTS) to enable retrieval of used nuclear fuel (UNF) for inspection or repackaging will increase as the duration and quantity of fuel in dry storage increases. This report explores the uses for a DTS, identifies associated general functional requirements, and reviews existing and proposed systems that currently perform dry fuel transfers. The focus of this paper is on the need for a DTS to enable transfer of bare fuel assemblies. Dry transfer systems for UNF canisters are currently available and in use for transferring loaded canisters between the drying station and storage and transportation casks.

  12. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  13. Off-normal performance of EBR-II (Experimental Breeder Reactor) driver fuel

    SciTech Connect

    Seidel, B.R.; Batte, G.L.; Lahm, C.E.; Fryer, R.M.; Koenig, J.F.; Hofman, G.L.

    1986-09-01

    The off-normal performance of EBR-II Mark-II driver fuel has been more than satisfactory as demonstrated by robust reliability under repeated transient overpower and undercooled loss-of-flow tests, by benign run-beyond-cladding-breach behavior, and by forgiving response to fabrication defects including lack of bond. Test results have verified that the metallic driver fuel is very tolerant of off-normal events. This behavior has allowed EBR-II to operate in a combined steady-state and transient mode to provide test capability without limitation from the metallic driver fuel.

  14. Nuclear fuel particles and method of making nuclear fuel compacts therefrom

    DOEpatents

    DeVelasco, Rubin I.; Adams, Charles C.

    1991-01-01

    Methods for making nuclear fuel compacts exhibiting low heavy metal contamination and fewer defective coatings following compact fabrication from a mixture of hardenable binder, such as petroleum pitch, and nuclear fuel particles having multiple layer fission-product-retentive coatings, with the dense outermost layer of the fission-product-retentive coating being surrounded by a protective overcoating, e.g., pyrocarbon having a density between about 1 and 1.3 g/cm.sup.3. Such particles can be pre-compacted in molds under relatively high pressures and then combined with a fluid binder which is ultimately carbonized to produce carbonaceous nuclear fuel compacts having relatively high fuel loadings.

  15. Waste Stream Analyses for Nuclear Fuel Cycles

    SciTech Connect

    N. R. Soelberg

    2010-08-01

    A high-level study was performed in Fiscal Year 2009 for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) Advanced Fuel Cycle Initiative (AFCI) to provide information for a range of nuclear fuel cycle options (Wigeland 2009). At that time, some fuel cycle options could not be adequately evaluated since they were not well defined and lacked sufficient information. As a result, five families of these fuel cycle options are being studied during Fiscal Year 2010 by the Systems Analysis Campaign for the DOE NE Fuel Cycle Research and Development (FCRD) program. The quality and completeness of data available to date for the fuel cycle options is insufficient to perform quantitative radioactive waste analyses using recommended metrics. This study has been limited thus far to qualitative analyses of waste streams from the candidate fuel cycle options, because quantitative data for wastes from the front end, fuel fabrication, reactor core structure, and used fuel for these options is generally not yet available.

  16. Irradiated Nuclear Fuel Management: Resource Versus Waste

    SciTech Connect

    Nash, Kenneth L.; Lumetta, Gregg J.; Vienna, John D.

    2013-01-01

    Management of irradiated fuel is an important component of commercial nuclear power production. Although it is broadly agreed that the disposition of some fraction of the fuel in geological repositories will be necessary, there is a range of options that can be considered that affect exactly what fraction of material will be disposed in that manner. Furthermore, until geological repositories are available to accept commercial irradiated fuel, these materials must be safely stored. Temporary storage of irradiated fuel has traditionally been conducted in storage pools, and this is still true for freshly discharged fuel. Criticality control technologies have led to greater efficiencies in packing of irradiated fuel into storage pools. With continued delays in establishing permanent repositories, utilities have begun to move some of the irradiated fuel inventory into dry storage. Fuel cycle options being considered worldwide include the once-through fuel cycle, limited recycle in which U and Pu are recycled back to power reactors as mixed oxide fuel, and advance partitioning and transmutation schemes designed to reduce the long term hazards associated with geological disposal from millions of years to a few hundred years. Each of these options introduces specific challenges in terms of the waste forms required to safely immobilize the hazardous components of irradiated fuel.

  17. MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

    DOEpatents

    Barton, C.J.; Grimes, W.R.

    1960-01-01

    Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

  18. Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

    ERIC Educational Resources Information Center

    Settle, Frank A.

    2009-01-01

    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 engineering of controlled fission are central to the generation of nuclear power, chemistry…

  19. International nuclear fuel cycle fact book

    SciTech Connect

    Leigh, I.W.

    1988-01-01

    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 or 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 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.

  20. International Nuclear Fuel Cycle Fact Book

    SciTech Connect

    Leigh, I W; Mitchell, S J

    1990-01-01

    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 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/NEA activities reports; proceedings of conferences and workshops, etc. The data listed do not reflect any one single source but frequently represent a consolidation/combination of information.

  1. International Nuclear Fuel Cycle Fact Book

    SciTech Connect

    Leigh, I.W.

    1992-05-01

    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.

  2. Advanced Nuclear Fuel Development in Japan

    NASA Astrophysics Data System (ADS)

    Yamawaki, Michio

    2003-06-01

    The verification test programs of high burnup BWR and PWR fuels have been carried out by Nuclear Power Engineering Corporation under the sponsorship of Ministry of Economy, Trade and Industry since 1986. BWR and PWR fuel assemblies of high burnup range of up to about 48 GWd/t and 53 GWd/t, respectively were examined by hot cell PIEs and many segment rods of local burnup range of up to more than 60GWd/t were power ramped in test reactors. Though some fuel rods showed minor failure after power ramp tests beyond commercial reactor condition, the results have shown good performance of the high burnup fuels in general. In BWR power ramp tests, the new failure mode of segment rods and the decrease of the failure threshold for higher burnup fuels have been found. Other than oxide fuel, new type fuels such as metallic, nitride and hydride fuels are under research and development in Japan for fast breeder reactors and, in case of hydride fuel, for both fast reactors and LWRs. Topics on some of these new type fuels will be also presented.

  3. Double-clad nuclear fuel safety rod

    DOEpatents

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    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.

  4. Spent nuclear fuel project integrated schedule plan

    SciTech Connect

    Squires, K.G.

    1995-03-06

    The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel.

  5. Computational Design of Advanced Nuclear Fuels

    SciTech Connect

    Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

    2014-06-03

    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.

  6. Nuclear fuel elements made from nanophase materials

    DOEpatents

    Heubeck, Norman B.

    1998-01-01

    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.degree. 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.

  7. Nuclear fuel elements made from nanophase materials

    DOEpatents

    Heubeck, N.B.

    1998-09-08

    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.

  8. Options contracts in the nuclear fuel industry

    SciTech Connect

    Fuller, D.M.

    1995-04-01

    This article discusses options trading in the nuclear fuels industry. Although there now exists no formal options market in the nuclear industry, flexibilities, or embedded options, are actually quite common in the long-term supply contracts. The value of these flexibilities can be estimated by applying the methods used to evaluate options. The method used is the Black-Scholes Model, and it is applied to a number of examples.

  9. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    DOEpatents

    Schluderberg, D.C.; Ryon, J.W.

    1962-05-01

    A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

  10. Nuclear fuel performance: Trends, remedies and challenges

    NASA Astrophysics Data System (ADS)

    Rusch, C. A.

    2008-12-01

    It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

  11. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom; Howard, Rob L; Flanagan, Michelle; Bjorkman, Gordon

    2014-01-01

    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.

  12. Nuclear Energy and Synthetic Liquid Transportation Fuels

    NASA Astrophysics Data System (ADS)

    McDonald, Richard

    2012-10-01

    This talk will propose a plan to combine nuclear reactors with the Fischer-Tropsch (F-T) process to produce synthetic carbon-neutral liquid transportation fuels from sea water. These fuels can be formed from the hydrogen and carbon dioxide in sea water and will burn to water and carbon dioxide in a cycle powered by nuclear reactors. The F-T process was developed nearly 100 years ago as a method of synthesizing liquid fuels from coal. This process presently provides commercial liquid fuels in South Africa, Malaysia, and Qatar, mainly using natural gas as a feedstock. Nuclear energy can be used to separate water into hydrogen and oxygen as well as to extract carbon dioxide from sea water using ion exchange technology. The carbon dioxide and hydrogen react to form synthesis gas, the mixture needed at the beginning of the F-T process. Following further refining, the products, typically diesel and Jet-A, can use existing infrastructure and can power conventional engines with little or no modification. We can then use these carbon-neutral liquid fuels conveniently long into the future with few adverse environmental impacts.

  13. Hazards of the Nuclear Fuel Cycle

    ERIC Educational Resources Information Center

    Holdren, John P.

    1974-01-01

    Outlines the stages of the nuclear fuel cycle where routine radiation releases occur and where nonroutine releases could occur. Examines the impact of these occurrences and emphasizes the regulations, practices, and technologies that prevail in the United States. (Author/GS)

  14. Spent Nuclear Fuel Alternative Technology Decision Analysis

    SciTech Connect

    Shedrow, C.B.

    1999-11-29

    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.

  15. Classical Molecular Dynamics Simulation of Nuclear Fuel

    SciTech Connect

    Devanathan, Ram; Krack, Matthias; Bertolus, Marjorie

    2015-10-10

    Molecular dynamics simulation is well suited to study primary damage production by irradiation, defect interactions with fission gas atoms, gas bubble nucleation, grain boundary effects on defect and gas bubble evolution in nuclear fuel, and the resulting changes in thermo-mechanical properties. In these simulations, the forces on the ions are dictated by interaction potentials generated by fitting properties of interest to experimental data. The results obtained from the present generation of potentials are qualitatively similar, but quantitatively different. There is a need to refine existing potentials to provide a better representation of the performance of polycrystalline fuel under a variety of operating conditions, and to develop models that are equipped to handle deviations from stoichiometry. In addition to providing insights into fundamental mechanisms governing the behaviour of nuclear fuel, MD simulations can also provide parameters that can be used as inputs for mesoscale models.

  16. Summary of nuclear fuel reprocessing activities around the world

    SciTech Connect

    Mellinger, P.J.; Harmon, K.M.; Lakey, L.T.

    1984-11-01

    This review of international practices for nuclear fuel reprocessing was prepared to provide a nontechnical summary of the current status of nuclear fuel reprocessing activities around the world. The sources of information are widely varied.

  17. Innovative nuclear fuels: results and strategy

    SciTech Connect

    Stan, Marius

    2009-01-01

    To facilitate the discovery and design of innovative nuclear fuels, multi-scale models and simulations are used to predict irradiation effects on the thermal conductivity, oxygen diffusivity, and thermal expansion of oxide fuels. The multi-scale approach is illustrated using results on ceramic fuels with a focus on predictions of point defect concentrations, stoichiometry, and phase stability. The high performance computer simulations include coupled heat transport, diffusion, and thermal expansion, gas bubble formation and temperature evolution in a fuel element consisting of UO2 fuel and metallic cladding. The second part of the talk is dedicated to a discussion of an international strategy for developing advanced, innovative nuclear fuels. Four initiative are proposed to accelerate the discovery and design of new materials: (a) Develop an international pool of experts, (b) Create Institutes for Materials Discovery and Design, (c) Create an International Knowledge base for experimental data, models (mathematical expressions), and simulations (codes) and (d) Organize international workshops and conference sessions. The paper ends with a discussion of existing and emerging international collaborations.

  18. Nuclear Fuel Cycle Options Catalog FY15 Improvements and Additions.

    SciTech Connect

    Price, Laura L.; Barela, Amanda Crystal; Schetnan, Richard Reed; Walkow, Walter M.

    2015-11-01

    The United States Department of Energy, Office of Nuclear Energy, Fuel Cycle Technology Program sponsors nuclear fuel cycle research and development. As part of its Fuel Cycle Options campaign, the DOE has established the Nuclear Fuel Cycle Options Catalog. The catalog is intended for use by the Fuel Cycle Technologies Program in planning its research and development activities and disseminating information regarding nuclear energy to interested parties. The purpose of this report is to document the improvements and additions that have been made to the Nuclear Fuel Cycle Options Catalog in the 2015 fiscal year.

  19. Sustainability Considerations in Spent Light-water Nuclear Fuel Retrievability

    SciTech Connect

    Wood, Thomas W.; Rothwell, Geoffrey

    2012-01-10

    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.

  20. Supply Security in Future Nuclear Fuel Markets

    SciTech Connect

    Seward, Amy M.; Wood, Thomas W.; Gitau, Ernest T.; Ford, Benjamin E.

    2013-11-18

    Previous PNNL work has shown the existing nuclear fuel markets to provide a high degree of supply security, including the ability to respond to supply disruptions that occur for technical and non-technical reasons. It is in the context of new reactor designs – that is, reactors likely to be licensed and market ready over the next several decades – that fuel supply security is most relevant. Whereas the fuel design and fabrication technology for existing reactors are well known, the construction of a new set of reactors could stress the ability of the existing market to provide adequate supply redundancy. This study shows this is unlikely to occur for at least thirty years, as most reactors likely to be built in the next three decades will be evolutions of current designs, with similar fuel designs to existing reactors.

  1. Storage assembly for spent nuclear fuel

    SciTech Connect

    Lapides, M.E.

    1982-04-27

    A technique for storing spent fuel rods from a nuclear reactor is disclosed herein. This technique utilizes a housing including a closed inner chamber for containing the fuel rods and a thermally conductive member located partially within the housing chamber and partially outside the housing for transferring heat generated by the fuel rods from the chamber to the ambient surroundings. Particulate material is located within the chamber and surrounds the fuel rods contained therein. This material is selected to serve as a heat transfer media between the contained cells and the heat transferring member and, at the same time, stand ready to fuse into a solid mass around the contained cells if the heat transferring member malfunctions or otherwise fails to transfer the generated heat out of the housing chamber in a predetermined way.

  2. Nuclear Fuel Leasing, Recycling, and Proliferation: Modeling a Global View

    SciTech Connect

    Reis, Victor H.; Crozat, Matthew P.; Choi, Jor-Shan; Hill, Robert

    2005-05-15

    A system dynamics model was created to simulate fuel cycle interactions between two separate nuclear entities, and this model was employed to investigate fuel leasing arrangements. The model was also adapted to evaluate proliferation and economic implications of an international leasing regime. For a nuclear growth scenario, an open fuel cycle results in extensive spent-fuel accumulation. For a closed fuel cycle, the leasing fuel cycle shows potential to reduce proliferation concern, especially if coupled with improved security and safeguard technology.

  3. Nuclear power generation and fuel cycle report 1996

    SciTech Connect

    1996-10-01

    This report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the worldwide nuclear fuel market. Long term projections of U.S. nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed. A discussion on decommissioning of nuclear power plants is included.

  4. Fuel cycle analysis of once-through nuclear systems.

    SciTech Connect

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

    2010-08-10

    (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

  5. Nuclear reactor fuel rod attachment system

    DOEpatents

    Not Available

    1980-09-17

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

  6. Method and apparatus for close packing of nuclear fuel assemblies

    DOEpatents

    Newman, Darrell F.

    1993-01-01

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  7. Method and apparatus for close packing of nuclear fuel assemblies

    DOEpatents

    Newman, D.F.

    1993-03-30

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  8. Hanford spent nuclear fuel project update

    SciTech Connect

    Williams, N.H.

    1997-08-19

    Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building.

  9. Nuclear rocket using indigenous Martian fuel NIMF

    NASA Technical Reports Server (NTRS)

    Zubrin, Robert

    1991-01-01

    In the 1960's, Nuclear Thermal Rocket (NTR) engines were developed and ground tested capable of yielding isp of up to 900 s at thrusts up to 250 klb. Numerous trade studies have shown that such traditional hydrogen fueled NTR engines can reduce the inertial mass low earth orbit (IMLEO) of lunar missions by 35 percent and Mars missions by 50 to 65 percent. The same personnel and facilities used to revive the hydrogen NTR can also be used to develop NTR engines capable of using indigenous Martian volatiles as propellant. By putting this capacity of the NTR to work in a Mars descent/acent vehicle, the Nuclear rocket using Indigenous Martian Fuel (NIMF) can greatly reduce the IMLEO of a manned Mars mission, while giving the mission unlimited planetwide mobility.

  10. Radiation protection at nuclear fuel cycle facilities.

    PubMed

    Endo, Kuniaki; Momose, Takumaro; Furuta, Sadaaki

    2011-07-01

    Radiation protection methodologies concerning individual monitoring, workplace monitoring and environmental monitoring in nuclear fuel facilities have been developed and applied to facilities in the Nuclear Fuel Cycle Engineering Laboratories (NCL) of Japan Atomic Energy Agency (JAEA) for over 40 y. External exposure to photon, beta ray and neutron and internal exposure to alpha emitter are important issues for radiation protection at these facilities. Monitoring of airborne and surface contamination by alpha and beta/photon emitters at workplace is also essential to avoid internal exposure. A critical accident alarm system developed by JAEA has been proved through application at the facilities for a long time. A centralised area monitoring system is effective for emergency situations. Air and liquid effluents from facilities are monitored by continuous monitors or sampling methods to comply with regulations. Effluent monitoring has been carried out for 40 y to assess the radiological impacts on the public and the environment due to plant operation.

  11. NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

    DOEpatents

    Rasor, N.S.; Hirsch, R.L.

    1963-12-01

    The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

  12. Holdup measurement for nuclear fuel manufacturing plants

    SciTech Connect

    Zucker, M.S.; Degen, M.; Cohen, I.; Gody, A.; Summers, R.; Bisset, P.; Shaub, E.; Holody, D.

    1981-07-13

    The assay of nuclear material holdup in fuel manufacturing plants is a laborious but often necessary part of completing the material balance. A range of instruments, standards, and a methodology for assaying holdup has been developed. The objectives of holdup measurement are ascertaining the amount, distribution, and how firmly fixed the SNM is. The purposes are reconciliation of material unbalance during or after a manufacturing campaign or plant decommissioning, to decide security requirements, or whether further recovery efforts are justified.

  13. Financing Strategies for Nuclear Fuel Cycle Facility

    SciTech Connect

    David Shropshire; Sharon Chandler

    2005-12-01

    To help meet our nation’s energy needs, reprocessing of spent nuclear fuel is being considered more and more as a necessary step in a future nuclear fuel cycle, but incorporating this step into the fuel cycle will require considerable investment. This report presents an evaluation of financing scenarios for reprocessing facilities integrated into the nuclear fuel cycle. A range of options, from fully government owned to fully private owned, was evaluated using a DPL (Dynamic Programming Language) 6.0 model, which can systematically optimize outcomes based on user-defined criteria (e.g., lowest life-cycle cost, lowest unit cost). Though all business decisions follow similar logic with regard to financing, reprocessing facilities are an exception due to the range of financing options available. The evaluation concludes that lowest unit costs and lifetime costs follow a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. Other financing arrangements, however, including regulated utility ownership and a hybrid ownership scheme, led to acceptable costs, below the Nuclear Energy Agency published estimates. Overwhelmingly, uncertainty in annual capacity led to the greatest fluctuations 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; the annual operating costs dominate the government case. It is concluded that to finance the construction and operation of such a facility without government ownership could be feasible with measures taken to mitigate risk, and that factors besides unit costs should be considered (e.g., legal issues, social effects, proliferation concerns) before making a decision on financing strategy.

  14. POWER GENERATION FROM LIQUID METAL NUCLEAR FUEL

    DOEpatents

    Dwyer, O.E.

    1958-12-23

    A nuclear reactor system is described wherein the reactor is the type using a liquid metal fuel, such as a dispersion of fissile material in bismuth. The reactor is designed ln the form of a closed loop having a core sectlon and heat exchanger sections. The liquid fuel is clrculated through the loop undergoing flssion in the core section to produce heat energy and transferrlng this heat energy to secondary fluids in the heat exchanger sections. The fission in the core may be produced by a separate neutron source or by a selfsustained chain reaction of the liquid fuel present in the core section. Additional auxiliary heat exchangers are used in the system to convert water into steam which drives a turbine.

  15. Refueling Behavior of Flexible Fuel Vehicle Drivers in the Federal Fleet

    SciTech Connect

    Daley, R.; Nangle, J.; Boeckman, G.; Miller, M.

    2014-05-01

    Federal fleets are a frequent subject of legislative and executive efforts to lead a national transition to alternative fuels and advanced vehicle technologies. Section 701 of the Energy Policy Act of 2005 requires that all dual-fueled alternative fuel vehicles in the federal fleet be operated on alternative fuel 100% of the time when they have access to it. However, in Fiscal Year (FY) 2012, drivers of federal flex fuel vehicles (FFV) leased through the General Services Administration refueled with E85 24% of the time when it was available--falling well short of the mandate. The U.S. Department of Energy's National Renewable Energy Laboratory completed a 2-year Laboratory Directed Research and Development project to identify the factors that influence the refueling behavior of federal FFV drivers. The project began with two primary hypotheses. First, information scarcity increases the tendency to miss opportunities to purchase E85. Second, even with perfect information, there are limits to how far drivers will go out of their way to purchase E85. This paper discusses the results of the project, which included a June 2012 survey of federal fleet drivers and an empirical analysis of actual refueling behavior from FY 2009 to 2012. This research will aid in the design and implementation of intervention programs aimed at increasing alternative fuel use and reducing petroleum consumption.

  16. Optimally moderated nuclear fission reactor and fuel source therefor

    DOEpatents

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  17. Transportation capabilities study of DOE-owned spent nuclear fuel

    SciTech Connect

    Clark, G.L.; Johnson, R.A.; Smith, R.W.; Abbott, D.G.; Tyacke, M.J.

    1994-10-01

    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.

  18. Dynamic Systems Analysis Report for Nuclear Fuel Recycle

    SciTech Connect

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

    2008-12-01

    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.

  19. EXTENDING SODIUM FAST REACTOR DRIVER FUEL USE TO HIGHER TEMPERATURES

    SciTech Connect

    Douglas L. Porter

    2011-02-01

    Calculations of potential sodium-cooled fast reactor fuel temperatures were performed to estimate the effects of increasing the outlet temperature of a given fast reactor design by increasing pin power, decreasing assembly flow, or increasing inlet temperature. Based upon experience in the U.S., both metal and mixed oxide (MOX) fuel types are discussed in terms of potential performance effects created by the increased operating temperatures. Assembly outlet temperatures of 600, 650 and 700 °C were used as goal temperatures. Fuel/cladding chemical interaction (FCCI) and fuel melting, as well as challenges to the mechanical integrity of the cladding material, were identified as the limiting phenomena. For example, starting with a recent 1000 MWth fast reactor design, raising the outlet temperature to 650 °C through pin power increase increased the MOX centerline temperature to more than 3300 °C and the metal fuel peak cladding temperature to more than 700 °C. These exceeded limitations to fuel performance; fuel melting was limiting for MOX and FCCI for metal fuel. Both could be alleviated by design ‘fixes’, such as using a barrier inside the cladding to minimize FCCI in the metal fuel, or using annular fuel in the case of MOX. Both would also require an advanced cladding material with improved stress rupture properties. While some of these are costly, the benefits of having a high-temperature reactor which can support hydrogen production, or other missions requiring high process heat may make the extra costs justified.

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

    SciTech Connect

    2013-07-01

    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.

  1. Pyroprocess for processing spent nuclear fuel

    DOEpatents

    Miller, William E.; Tomczuk, Zygmunt

    2002-01-01

    This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF.sub.4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF.sub.4 with a eutectic melting point of 500.degree. C. Prior to lowering the basket, the salt is heated to a temperature of between 550.degree. C. and 700.degree. C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF.sub.6. In addition, after dissolution, the basket contains PuO.sub.2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

  2. Spent Nuclear Fuel Alternative Technology Risk Assessment

    SciTech Connect

    Perella, V.F.

    1999-11-29

    A Research Reactor Spent Nuclear Fuel Task Team (RRTT) was chartered by the Department of Energy (DOE) Office of Spent Fuel Management with the responsibility to recommend a course of action leading to a final technology selection for the interim management and ultimate disposition of the foreign and domestic aluminum-based research reactor spent nuclear fuel (SNF) under DOE''s jurisdiction. The RRTT evaluated eleven potential SNF management technologies and recommended that two technologies, direct co-disposal and an isotopic dilution alternative, either press and dilute or melt and dilute, be developed in parallel. Based upon that recommendation, the Westinghouse Savannah River Company (WSRC) organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and provide a WSRC recommendation to DOE for a preferred SNF alternative management technology. A technology risk assessment was conducted as a first step in this recommendation process to determine if either, or both, of the technologies posed significant risks that would make them unsuitable for further development. This report provides the results of that technology risk assessment.

  3. Spent Nuclear Fuel Vibration Integrity Study

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Jiang, Hao; Yan, Yong; Bevard, Bruce Balkcom

    2016-01-01

    The objective of this research is to collect dynamic experimental data on spent nuclear fuel (SNF) under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT), the hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The collected CIRFT data will be utilized to support ongoing spent fuel modeling activities, and support SNF transportation related licensing issues. Recent testing to understand the effects of hydride reorientation on SNF vibration integrity is also being evaluated. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance, Fuel structure contributes to the SNF system stiffness, There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response, and that data will also be presented.

  4. Nuclear reactor fuel rod attachment system

    DOEpatents

    Christiansen, David W.

    1982-01-01

    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.

  5. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-01-01

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  6. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-12-02

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  7. Nuclear fuel for liquid metal cooled nuclear reactors

    SciTech Connect

    Duncombe, E.; Adamson, J.; Gratton, C.P.

    1983-11-22

    In a cluster of nuclear fuel rods cooled by liquid metal an obstruction to coolant flow results in overheating in the wake of the obstruction. By the provision of open ended heat transfer tubes in the flow channels, a guaranteed supply of coolant is maintained and this supply holds the temperature to below saturation. Heat transfer via the tubes is highly efficient and ensures that a sufficient temperature rise occurs at the cluster exit to provoke a response from the outlet temperature transducer sensing average temperature.

  8. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    SciTech Connect

    Jubin, R. T.; Strachan, D. M.; Ilas, G.; Spencer, B. B.; Soelberg, N. R.

    2014-09-01

    In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14C, 3H, 129I, and 85Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is

  9. Energy Fuels Nuclear, Inc. Arizona Strip Operations

    SciTech Connect

    Pool, T.C.

    1993-05-01

    Founded in 1975 by uranium pioneer, Robert W. Adams, Energy Fuels Nuclear, Inc. (EFNI) emerged as the largest US uranium mining company by the mid-1980s. Confronting the challenges of declining uranium market prices and the development of high-grade ore bodies in Australia and Canada, EFNI aggressively pursued exploration and development of breccia-pipe ore bodies in Northwestern Arizona. As a result, EFNI's production for the Arizona Strip of 18.9 million pounds U[sub 3]O[sub 8] over the period 1980 through 1991, maintained the company's status as a leading US uranium producer.

  10. International nuclear fuel cycle fact book. Revision 6

    SciTech Connect

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

    1986-01-01

    The International Fuel Cycle 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. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2.

  11. Electrorefining Experience For Pyrochemical Reprocessing of Spent EBR-II Driver Fuel

    SciTech Connect

    S. X. Li; T. A. Johnson; B. R. Westphal; K. M. Goff; R. W. Benedict

    2005-10-01

    Pyrochemical processing has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor-II (EBR-II) at Idaho National Laboratory since 1996. This report summarizes technical advancements made in electrorefining of spent EBR-II driver fuel in the Mk-IV electrorefiner since the pyrochemical processing was integrated into the AFCI program in 2002. The significant advancements include improving uranium dissolution and noble metal retention from chopped fuel segments, increasing cathode current efficiency, and achieving co-collection of zirconium along with uranium from the cadmium pool.

  12. Method and means of packaging nuclear fuel rods for handling

    DOEpatents

    Adam, Milton F.

    1979-01-01

    Nuclear fuel rods, especially spent nuclear fuel rods that may show physical distortion, are encased within a metallic enclosing structure by forming a tube about the fuel rod. The tube has previously been rolled to form an overlapping tubular structure and then unrolled and coiled about an axis perpendicular to the tube. The fuel rod is inserted into the tube as the rolled tube is removed from a coiled strip and allowed to reassume its tubular shape about the fuel rod. Rollers support the coiled strip in an open position as the coiled strip is uncoiled and allowed to roll about the fuel rod.

  13. Characterization of Hydrogen Content in ZIRCALOY-4 Nuclear Fuel Cladding

    NASA Astrophysics Data System (ADS)

    Pfeif, E. A.; Lasseigne, A. N.; Krzywosz, K.; Mader, E. V.; Mishra, B.; Olson, D. L.

    2010-02-01

    Assessment of hydrogen uptake of underwater nuclear fuel clad and component materials will enable improved monitoring of fuel health. Zirconium alloys are used in nuclear reactors as fuel cladding, fuel channels, guide tubes and spacer grids, and are available for inspection in spent fuel pools. With increasing reactor exposure zirconium alloys experience hydrogen ingress due to neutron interactions and water-side corrosion that is not easily quantified without destructive hot cell examination. Contact and non-contact nondestructive techniques, using Seebeck coefficient measurements and low frequency impedance spectroscopy, to assess the hydrogen content and hydride formation within zircaloy 4 material that are submerged to simulate spent fuel pools are presented.

  14. Separator assembly for use in spent nuclear fuel shipping cask

    DOEpatents

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  15. Fuel element concept for long life high power nuclear reactors

    NASA Technical Reports Server (NTRS)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  16. Characterization plan for Hanford spent nuclear fuel

    SciTech Connect

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

    1994-12-01

    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.

  17. Electric heater for nuclear fuel rod simulators

    DOEpatents

    McCulloch, Reginald W.; Morgan, Jr., Chester S.; Dial, Ralph E.

    1982-01-01

    The present invention is directed to an electric cartridge-type heater for use as a simulator for a nuclear fuel pin in reactor studies. The heater comprises an elongated cylindrical housing containing a longitudinally extending helically wound heating element with the heating element radially inwardly separated from the housing. Crushed cold-pressed preforms of boron nitride electrically insulate the heating element from the housing while providing good thermal conductivity. Crushed cold-pressed preforms of magnesia or a magnesia-15 percent boron nitride mixture are disposed in the cavity of the helical heating element. The coefficient of thermal expansion of the magnesia or the magnesia-boron nitride mixture is higher than that of the boron nitride disposed about the heating element for urging the boron nitride radially outwardly against the housing during elevated temperatures to assure adequate thermal contact between the housing and the boron nitride.

  18. Nevada commercial spent nuclear fuel transportation experience

    SciTech Connect

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed.

  19. Spent nuclear fuel project technical databook

    SciTech Connect

    Reilly, M.A.

    1998-07-22

    The Spent Nuclear Fuel (SNF) project technical databook provides project-approved summary tables of selected parameters and derived physical quantities, with nominal design and safety basis values. It contains the parameters necessary for a complete documentation basis of the SNF Project technical and safety baseline. The databook is presented in two volumes. Volume 1 presents K Basins SNF related information. Volume 2 (not yet available) will present selected sludge and water information, as it relates to the sludge and water removal projects. The values, within this databook, shall be used as the foundation for analyses, modeling, assumptions, or other input to SNF project safety analyses or design. All analysis and modeling using a parameter available in this databook are required to use and cite the appropriate associated value, and document any changes to those values (i.e., analysis assumptions, equipment conditions, etc). Characterization and analysis efforts are ongoing to validate, or update these values.

  20. Nuclear Fuel Plate Fabrication Employing Friction Welding

    SciTech Connect

    Douglas E. Burkes; Neil P. Hallinan; Curtis R. Clark

    2008-09-01

    This paper provides an overview of the friction bonding process, a novel modification of the more conventional friction stir welding process. The process has been modified to enable the fabrication of plate-type nuclear fuels for the conversion of research and test reactors currently operating using highly enriched uranium to low-enriched uranium. Discussions related to the specific modifications of the friction bonding process have been provided, in addition to challenges associated with these modifications. Progression of the process and solutions to the challenges are provided so that users of the friction stir welding process and those investigating fabrication of other laminar composites, where joining of one or more layers is essential, might draw from the authors’ experiences discussed in this paper.

  1. Nuclear fuel post-irradiation examination equipment package

    SciTech Connect

    DeCooman, W.J.; Spellman, D.J.

    2007-07-01

    Hot cell capabilities in the U.S. are being reviewed and revived to meet today's demand for fuel reliability, tomorrow's demands for higher burnup fuel and future demand for fuel recycling. Fuel reliability, zero tolerance for failure, is more than an industry buzz. It is becoming a requirement to meet the rapidly escalating demands for the impending renaissance of nuclear power generation, fuel development, and management of new waste forms that will need to be dealt with from programs such as the Global Nuclear Energy Partnership (GNEP). Fuel performance data is required to license fuel for higher burnup; to verify recycled fuel performance, such as MOX, for wide-scale use in commercial reactors; and, possibly, to license fuel for a new generation of fast reactors. Additionally, fuel isotopic analysis and recycling technologies will be critical factors in the goal to eventually close the fuel cycle. This focus on fuel reliability coupled with the renewed interest in recycling puts a major spotlight on existing hot cell capabilities in the U.S. and their ability to provide the baseline analysis to achieve a closed fuel cycle. Hot cell examination equipment is necessary to determine the characteristics and performance of irradiated materials that are subjected to nuclear reactor environments. The equipment within the hot cells is typically operated via master-slave manipulators and is typically manually operated. The Oak Ridge National Laboratory is modernizing their hot cell nuclear fuel examination equipment, installing automated examination equipment and data gathering capabilities. Currently, the equipment has the capability to perform fuel rod visual examinations, length and diametrical measurements, eddy current examination, profilometry, gamma scanning, fission gas collection and void fraction measurement, and fuel rod segmentation. The used fuel postirradiation examination equipment was designed to examine full-length fuel rods for both Boiling Water

  2. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    SciTech Connect

    Quiter, Brian; Ludewigt, Bernhard; Ambers, Scott

    2011-06-30

    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

  3. Technology Readiness Levels for Advanced Nuclear Fuels and Materials Development

    SciTech Connect

    Jon Carmack

    2014-01-01

    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).

  4. Coupon Surveillance For Corrosion Monitoring In Nuclear Fuel Basin

    SciTech Connect

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

    2012-10-01

    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.

  5. World nuclear capacity and fuel cycle requirements, November 1993

    SciTech Connect

    Not Available

    1993-11-30

    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.

  6. Grooved Fuel Rings for Nuclear Thermal Rocket Engines

    NASA Technical Reports Server (NTRS)

    Emrich, William

    2009-01-01

    An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

  7. Accelerated Closure of the Spent Nuclear Fuel (SNF) project

    SciTech Connect

    RUTHERFORD, W.W.

    2001-02-01

    The K East and K West Basins, built in the early 1950s, have been used to store irradiated nuclear fuel from the Hanford N Reactor. This fuel, which is referred to as spent nuclear fuel (SNF), has been stored underwater since 1975 in KE Basin and since 1981 in KW Basin. There are 54,000 N Reactor fuel assemblies in 3,800 canisters in the K West Basin, and 51,000 fuel assemblies in 3,700 canisters in the K East Basin that total 2,100 metric tons of SNF.

  8. Spent nuclear fuel discharges from U.S. reactors 1994

    SciTech Connect

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  9. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect

    1998-03-01

    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.

  10. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2008-01-20

    reprocessing is currently conducted in France, Britain, and Russia. The 239Pu they produce is blended with uranium to make mixed- oxide ( MOX ) fuel, in which...Association, Mixed Oxide Fuel ( MOX ), November 2006, at [http://www.world-nuclear.org/info/inf29.html]. 53 World Nuclear Association, Uranium Markets, March...older plants to reprocess gas-cooled reactor fuel, and India has a 275-ton plant.51 About 200 metric tons of MOX fuel is used annually, about 2% of

  11. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2008-09-03

    http://www.world-nuclear.org/info/inf69.html]. 53 World Nuclear Association, Mixed Oxide Fuel ( MOX ), November 2006, at [http://www.world-nuclear.org...oxide ( MOX ) fuel, in which the 239Pu largely substitutes for 235U. Two French reprocessing plants at La Hague can each reprocess up to 800 metric tons of...to reprocess gas-cooled reactor fuel, and India has a 275-ton plant.52 About 200 metric tons of MOX fuel is used annually, about 2% of new nuclear

  12. Fuel rod retention device for a nuclear reactor

    DOEpatents

    Hylton, Charles L.

    1984-01-01

    A device is described for supporting a nuclear fuel rod in a fuel rod assembly which allows the rod to be removed without disturbing other rods in the assembly. A fuel rod cap connects the rod to a bolt which is supported in the assembly end fitting by means of a locking assembly. The device is designed so that the bolt is held securely during normal reactor operation yet may be easily disengaged and the fuel rod removed when desired.

  13. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    SciTech Connect

    Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.

    2013-09-30

    This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). 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 (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

  14. Methods for making a porous nuclear fuel element

    DOEpatents

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

    2014-12-30

    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.

  15. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect

    Not Available

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  16. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2009-07-01

    uranium to make mixed-oxide ( MOX ) fuel, in which the 239Pu largely substitutes for 235U. Two French reprocessing plants at La Hague can each...Managing the Nuclear Fuel Cycle Congressional Research Service 17 plant.54 About 200 metric tons of MOX fuel is used annually, about 2% of new nuclear...fuel,55 equivalent to about 2,000 metric tons of mined uranium.56 However, the benefits of reprocessing spent fuel to make MOX fuel for today’s

  17. Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement

    SciTech Connect

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

    2013-05-01

    A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

  18. Nuclear fuel corrosion over millennia interpreted using geologic data

    SciTech Connect

    Pearcy, E.C.; Manaktala, H.K.

    1994-12-31

    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.

  19. AN ANALYTICAL FRAMEWORK FOR ASSESSING RELIABLE NUCLEAR FUEL SERVICE APPROACHES: ECONOMIC AND NON-PROLIFERATION MERITS OF NUCLEAR FUEL LEASING

    SciTech Connect

    Kreyling, Sean J.; Brothers, Alan J.; Short, Steven M.; Phillips, Jon R.; Weimar, Mark R.

    2010-08-11

    The goal of international nuclear policy since the dawn of nuclear power has been the peaceful expansion of nuclear energy while controlling the spread of enrichment and reprocessing technology. Numerous initiatives undertaken in the intervening decades to develop international agreements on providing nuclear fuel supply assurances, or reliable nuclear fuel services (RNFS) attempted to control the spread of sensitive nuclear materials and technology. In order to inform the international debate and the development of government policy, PNNL has been developing an analytical framework to holistically evaluate the economics and non-proliferation merits of alternative approaches to managing the nuclear fuel cycle (i.e., cradle-to-grave). This paper provides an overview of the analytical framework and discusses preliminary results of an economic assessment of one RNFS approach: full-service nuclear fuel leasing. The specific focus of this paper is the metrics under development to systematically evaluate the non-proliferation merits of fuel-cycle management alternatives. Also discussed is the utility of an integrated assessment of the economics and non-proliferation merits of nuclear fuel leasing.

  20. Nuclear Islands: International Leasing of Nuclear Fuel Cycle Sites to Provide Enduring Assurance of Peaceful Use

    SciTech Connect

    Paine, Christopher E.; Cochran, Thomas B.

    2010-11-01

    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 reprocessing plants and other sensitive nuclear fuel cycle facilities. The INFCA would provide significantly increased nonproliferation assurance to its member states and the wider international community by holding long-term leasehold contracts to operate secure restricted zones containing such sensitive nuclear facilities.

  1. Evaluation of Waste Arising from Future Nuclear Fuel Cycle

    SciTech Connect

    Jubin, Robert Thomas; Taiwo, Temitope; Wigeland, Roald

    2015-01-01

    A comprehensive study was recently completed at the request of the US Department of Energy Office of Nuclear Energy (DOE-NE) to evaluate and screen nuclear fuel cycles. The final report was issued in October 2014. Uranium- and thorium-based fuel cycles were evaluated using both fast and thermal spectrum reactors. Once-through, limited-recycle, and continuous-recycle cases were considered. This study used nine evaluation criteria to identify promising fuel cycles. Nuclear waste management was one of the nine evaluation criteria. The waste generation criterion from this study is discussed herein.

  2. Laser-Based Characterization of Nuclear Fuel Plates

    SciTech Connect

    James A. Smith; David L. Cottle; Barry H. Rabin

    2013-07-01

    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.

  3. Laser-based characterization of nuclear fuel plates

    SciTech Connect

    Smith, James A.; Cottle, Dave L.; Rabin, Barry H.

    2014-02-18

    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.

  4. Pyroprocessing of fast flux test facility nuclear fuel

    SciTech Connect

    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.

    2013-07-01

    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)

  5. Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

    SciTech Connect

    B.R. Westphal; G.L. Fredrickson; G.G. Galbreth; D. Vaden; M.D. Elliott; J.C. Price; E.M. Honeyfield; M.N. Patterson; L. A. Wurth

    2013-10-01

    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 electrorefined uranium products exceeded 99%.

  6. Nuclear fuel alloys or mixtures and method of making thereof

    DOEpatents

    Mariani, Robert Dominick; Porter, Douglas Lloyd

    2016-04-05

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

  7. FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Abbott, W.E.; Balent, R.

    1958-09-16

    A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

  8. Radiation dose aspects in the handling of emerging nuclear fuels.

    PubMed

    Nicolaou, G

    2014-12-01

    The occupational annual dose levels, encountered at fabrication of emerging nuclear fuels, have been studied. Emerging fuels for the single and multiple recycling of Pu and MA have resulted in considerably higher gamma and neutron doses in comparison with commercial fuels. The occupational dose limit is exceeded at fabrication by a single fuel rod in all fuel cases with (241)Am and Cm isotopes present in their composition. In the absence of these isotopes, 2-4 adjacent fuel rods are sufficient to exceed the limit. Self-shielding within the fuel reduces significantly only the gamma dose that would have been delivered otherwise. Hence, only the first row of fuel rods in an assembly contributes to the dose, whereas in the case of neutrons, all fuel rods contribute.

  9. Realistic Challenges to Optimize the Closed Nuclear Fuel Cycle

    SciTech Connect

    Shigeo Nomura; Hisao Ojima; Hirohisa Ishikawa

    2002-07-01

    The first generation nuclear fuel cycle system will be closed in Japan by 2010 with the successful combination of LWR-MOX, industrial operation of reprocessing and MOX fuel fabrication plants, interim storage of spent fuel, and plans for the location of final disposal of radioactive wastes. To tailor nuclear power as the globally sustainable and environmentally benign energy source for future society, however, an advanced fuel cycle option is needed that is well suited to the fast reactor. It should recycle fissile materials of plutonium and uranium whilst transmuting minor actinides and generating minimum waste. (authors)

  10. Method of increasing the deterrent to proliferation of nuclear fuels

    SciTech Connect

    Rampolla, D.S.

    1982-08-17

    A process is claimed of recycling protactinium-231 to enhance the utilization of radioactively hot uranium-232 in nuclear fuel for the purpose of making both fresh and spent fuel more resistant to proliferation. The uranium-232 may be obtained by the irradiation of protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor. The production of protactinium-231 and uranium-232 would be made possible by the use of the thorium uranium-233 fuel cycle in power reactors.

  11. Method of increasing the deterrent to proliferation of nuclear fuels

    DOEpatents

    Rampolla, Donald S.

    1982-01-01

    A process of recycling protactinium-231 to enhance the utilization of radioactively hot uranium-232 in nuclear fuel for the purpose of making both fresh and spent fuel more resistant to proliferation. The uranium-232 may be obtained by the irradiation of protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor. The production of protactinium-231 and uranium-232 would be made possible by the use of the thorium uranium-233 fuel cycle in power reactors.

  12. A Comparison Study of Various Nuclear Fuel Cycle Alternatives

    SciTech Connect

    Kwon, Eun-ha; Ko, Won-il

    2007-07-01

    As a nation develops its nuclear strategies, it must consider various aspects of nuclear energy such as sustainability, environmental-friendliness, proliferation-resistance, economics, technologies, and so on. Like all the policy decision, however, a nuclear fuel cycle option can not be superior in all aspects; the nation must identify its top priority and accordingly evaluate all the possible nuclear fuel cycle options. For such a purpose, this paper takes four different fuel cycle options that are likely adopted by the Korean government, considering the current status of nuclear power generation and the 3. Comprehensive Nuclear Energy Promotion Plan (CNEPP) - Once-through Cycle, DUPIC Recycle, Thermal Recycle and GEN-IV Recycle. The paper then evaluates each option in terms of resource utilization and waste generation. The analysis shows that the GEN-IV Recycle appears to be most competitive from these aspects. (authors)

  13. Fabrication of high exposure nuclear fuel pellets

    DOEpatents

    Frederickson, James R.

    1987-01-01

    A method is disclosed for making a fuel pellet for a nuclear reactor. A mixture is prepared of PuO.sub.2 and UO.sub.2 powders, where the mixture contains at least about 30% PuO.sub.2, and where at least about 12% of the Pu is the Pu.sup.240 isotope. To this mixture is added about 0.3 to about 5% of a binder having a melting point of at least about 250.degree. F. The mixture is pressed to form a slug and the slug is granulated. Up to about 4.7% of a lubricant having a melting point of at least about 330.degree. F. is added to the granulated slug. Both the binder and the lubricant are selected from a group consisting of polyvinyl carboxylate, polyvinyl alcohol, naturally occurring high molecular weight cellulosic polymers, chemically modified high molecular weight cellulosic polymers, and mixtures thereof. The mixture is pressed to form a pellet and the pellet is sintered.

  14. The dynamics and drivers of fuel and fire in the Portuguese public forest.

    PubMed

    Fernandes, Paulo M; Loureiro, Carlos; Guiomar, Nuno; Pezzatti, Gianni B; Manso, Filipa T; Lopes, Luís

    2014-12-15

    The assumption that increased wildfire incidence in the Mediterranean Basin during the last decades is an outcome of changes in land use warrants an objective analysis. In this study we examine how annual area burned (BA) in the Portuguese public forest varied in relation to environmental and human-influenced drivers during the 1943-2011 period. Fire behaviour models were used to describe fuel hazard considering biomass removal, cover type changes, area burned, post-disturbance fuel accumulation, forest age-classes distribution and fuel connectivity. Biomass removal decreased rapidly beyond the 1940s, which, along with afforestation, increased fuel hazard until the 1980s; a subsequent decline was caused by increased fire activity. Change point analysis indicates upward shifts in BA in 1952 and in 1973, both corresponding to six-fold increases. Fire weather (expressed by the 90th percentile of the Canadian FWI during summer) increased over the study period, accounting for 18 and 36% of log(BA) variation before 1974 and after 1973, respectively. Regression modelling indicates that BA responds positively to fire weather, fuel hazard and number of fires in descending order of importance; pre-summer and 2-year lagged precipitation respectively decrease and increase BA, but the effects are minor and non-significant when both variables are included in the model. Land use conflicts (expressed through more fires) played a role, but it was afforestation and agricultural abandonment that supported the fire regime shifts, explaining weather-drought as the current major driver of BA as well. We conclude that bottom-up factors, i.e. human-induced changes in landscape flammability and ignition density, can enhance or override the influence of weather-drought on the fire regime in Mediterranean humid regions. A more relevant role of fuel control in fire management policies and practices is warranted by our findings.

  15. Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory

    SciTech Connect

    Craft, Aaron E.; Wachs, Daniel M.; Okuniewski, Maria A.; Chichester, David L.; Williams, Walter J.; Papaioannou, Glen C.; Smolinski, Andrew T.

    2015-09-10

    Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This study describes the NRAD and hot cell facilities, the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging.

  16. Neutron Radiography of Irradiated Nuclear Fuel at Idaho National Laboratory

    NASA Astrophysics Data System (ADS)

    Craft, Aaron E.; Wachs, Daniel M.; Okuniewski, Maria A.; Chichester, David L.; Williams, Walter J.; Papaioannou, Glen C.; Smolinski, Andrew T.

    Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This paper describes the NRAD and hot cell facilities, the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging.

  17. Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory

    DOE PAGES

    Craft, Aaron E.; Wachs, Daniel M.; Okuniewski, Maria A.; ...

    2015-09-10

    Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This study describes the NRAD and hot cell facilities,more » the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging.« less

  18. Improved nuclear fuel assembly grid spacer

    DOEpatents

    Marshall, John; Kaplan, Samuel

    1977-01-01

    An improved fuel assembly grid spacer and method of retaining the basic fuel rod support elements in position within the fuel assembly containment channel. The improvement involves attachment of the grids to the hexagonal channel and of forming the basic fuel rod support element into a grid structure, which provides a design which is insensitive to potential channel distortion (ballooning) at high fluence levels. In addition the improved method eliminates problems associated with component fabrication and assembly.

  19. Energy Return on Investment from Recycling Nuclear Fuel

    SciTech Connect

    2011-08-17

    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.

  20. Nuclear reactor fuel element having improved heat transfer

    DOEpatents

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  1. Means for supporting fuel elements in a nuclear reactor

    DOEpatents

    Andrews, Harry N.; Keller, Herbert W.

    1980-01-01

    A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

  2. Coolant mass flow equalizer for nuclear fuel

    DOEpatents

    Betten, Paul R.

    1978-01-01

    The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

  3. Exploration for fossil and nuclear fuels from orbital altitudes

    NASA Technical Reports Server (NTRS)

    Short, N. M.

    1977-01-01

    The paper discusses the application of remotely sensed data from orbital satellites to the exploration for fossil and nuclear fuels. Geological applications of Landsat data are described including map editing, lithologic identification, structural geology, and mineral exploration. Specific results in fuel exploration are reviewed and a series of related Landsat images is included.

  4. Hanford`s spent nuclear fuel retrieval: an agressive agenda

    SciTech Connect

    Shen, E.J., Westinghouse Hanford

    1996-12-06

    Starting December 1997, spent nuclear fuel that has been stored in the K Reactor Fuel Storage Basins will be retrieved over a two year period and repackaged for long term dry storage. The aging and sometimes corroding fuel elements will be recovered and processed using log handled tools and teleoperated manipulator technology. The U.S. Department of Energy (DOE) is committed to this urgent schedule because of the environmental threats to the groundwater and nearby the Columbia River.

  5. Support grid for fuel elements in a nuclear reactor

    DOEpatents

    Finch, Lester M.

    1977-01-01

    A support grid is provided for holding nuclear fuel rods in a rectangular array. Intersecting sheet metal strips are interconnected using opposing slots in the strips to form a rectangular cellular grid structure for engaging the sides of a multiplicity of fuel rods. Spring and dimple supports for engaging fuel and guide rods extending through each cell in the support grid are formed in the metal strips with the springs thus formed being characterized by nonlinear spring rates.

  6. Advanced nuclear fuel cycles - Main challenges and strategic choices

    SciTech Connect

    Le Biez, V.; Machiels, A.; Sowder, A.

    2013-07-01

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness.

  7. Reprocessing of nuclear fuels at the Savannah River Plant

    SciTech Connect

    Gray, L.W.

    1986-10-04

    For more than 30 years, the Savannah River Plant (SRP) has been a major supplier of nuclear materials such as plutonium-239 and tritium-3 for nuclear and thermonuclear weapons, plutonium-238 for space exploration, and isotopes of americium, curium, and californium for use in the nuclear research community. SRP is a complete nuclear park, providing most of the processes in the nuclear fuel cycle. Key processes involve fabrication and cladding of the nuclear fuel, target, and control assemblies; rework of heavy water for use as reactor moderator; reactor loading, operation, and unloading; chemical recovery of the reactor transmutation products and spent fuels; and management of the gaseous, liquid, and solid nuclear and chemical wastes; plus a host of support operations. The site's history and the key processes from fabrication of reactor fuels and targets to finishing of virgin plutonium for use in the nuclear weapons complex are reviewed. Emphasis has been given to the chemistry of the recovery and purification of weapons grade plutonium from irradiated reactor targets.

  8. Mox fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-05-15

    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.

  9. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    1998-01-01

    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.

  10. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-07-17

    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.

  11. MOX fuel arrangement for nuclear core

    DOEpatents

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

    1998-10-13

    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.

  12. Electrochemical fluorination for processing of used nuclear fuel

    DOEpatents

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2016-07-05

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  13. Galvanic cell for processing of used nuclear fuel

    DOEpatents

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2017-02-07

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  14. Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

    NASA Technical Reports Server (NTRS)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

    2012-01-01

    Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes thses overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

  15. Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

    NASA Technical Reports Server (NTRS)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

    2012-01-01

    Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes these overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

  16. K Basin spent nuclear fuel characterization

    SciTech Connect

    LAWRENCE, L.A.

    1999-02-10

    The results of the characterization efforts completed for the N Reactor fuel stored in the Hanford K Basins were Collected and summarized in this single referencable document. This summary provides a ''road map'' for what was done and the results obtained for the fuel characterization program initiated in 1994 and scheduled for completion in 1999 with the fuel oxidation rate measurement under moist inert atmospheres.

  17. Nuclear Forensics and Attribution for Improved Energy Security: The Use of Taggants in Nuclear Fuel

    SciTech Connect

    Kristo, M J; Robel, M; Hutcheon, I D

    2007-04-05

    The Global Nuclear Energy Partnership (GNEP), recently announced by DOE Secretary Bodman, poses significant new challenges with regard to securing, safeguarding, monitoring and tracking nuclear materials. In order to reduce the risk of nuclear proliferation, new technologies must be developed to reduce the risk that nuclear material can be diverted from its intended use. Regardless of the specific nature of the fuel cycle, nuclear forensics and attribution will play key roles to ensure the effectiveness of nonproliferation controls and to deter the likelihood of illicit activities. As the leader of the DHS nuclear and radiological pre-detonation attribution program, LLNL is uniquely positioned to play a national leadership role in this effort. Ensuring that individuals or organizations engaged in illicit trafficking are rapidly identified and apprehended following theft or diversion of nuclear material provides a strong deterrent against unlawful activities. Key to establishing this deterrent is developing the ability to rapidly and accurately determine the identity, source and prior use history of any interdicted nuclear material. Taggants offer one potentially effective means for positively identifying lost or stolen nuclear fuels. Taggants are materials that can be encoded with a unique signature and introduced into nuclear fuel during fuel fabrication. During a nuclear forensics investigation, the taggant signature can be recovered and the nuclear material identified through comparison with information stored in an appropriate database. Unlike serial numbers or barcodes, microtaggants can provide positive identification with only partial recovery, providing extreme resistance to any attempt to delete or alter them.

  18. Subcritical transmutation of spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Sommer, Christopher M.

    2011-07-01

    A series of fuel cycle simulations were performed using CEA's reactor physics code ERANOS 2.0 to analyze the transmutation performance of the Subcritical Advanced Burner Reactor (SABR). SABR is a fusion-fission hybrid reactor that combines the leading sodium cooled fast reactor technology with the leading tokamak plasma technology based on ITER physics. Two general fuel cycles were considered for the SABR system. The first fuel cycle is one in which all of the transuranics from light water reactors are burned in SABR. The second fuel cycle is a minor actinide burning fuel cycle in which all of the minor actinides and some of the plutonium produced in light water reactors are burned in SABR, with the excess plutonium being set aside for starting up fast reactors in the future. The minor actinide burning fuel cycle is being considered in European Scenario Studies. The fuel cycles were evaluated on the basis of TRU/MA transmutation rate, power profile, accumulated radiation damage, and decay heat to the repository. Each of the fuel cycles are compared against each other, and the minor actinide burning fuel cycles are compared against the EFIT transmutation system, and a low conversion ratio fast reactor.

  19. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    DOEpatents

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

    2010-02-23

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). 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, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

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

    DOEpatents

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

    2013-09-03

    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.

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

    DOEpatents

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

    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.

  2. International nuclear fuel cycle fact book. Revision 4

    SciTech Connect

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

    1984-03-01

    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.

  3. International Nuclear Fuel Cycle Fact Book. Revision 5

    SciTech Connect

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

    1985-01-01

    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.

  4. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    SciTech Connect

    P.M. O'Leary; Dr. M.L. Pitts

    2000-08-21

    Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers.

  5. Recapturing NERVA-Derived Fuels for Nuclear Thermal Propulsion

    SciTech Connect

    Qualls, A L; Hancock, Emily F

    2011-01-01

    The Department of Energy is working with NASA to examine fuel options for Nuclear Thermal Propulsion applications. Extensive development and testing was performed on graphite-based fuels during the Nuclear Engineer Rocket Vehicle Application (NERVA) and Rover programs through the early 1970s. This paper explores the possibility of recapturing the technology and the issues associated with using it for the next generation of nuclear thermal rockets. The issues discussed include a comparison of today's testing capabilities, analysis techniques and methods, and knowledge to that of previous development programs and presents a plan to recapture the technology for a flight program.

  6. Spent nuclear fuel discharges from US reactors 1992

    SciTech Connect

    Not Available

    1994-05-05

    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.

  7. Depleted uranium as a backfill for nuclear fuel waste package

    DOEpatents

    Forsberg, Charles W.

    1998-01-01

    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.

  8. Depleted uranium as a backfill for nuclear fuel waste package

    DOEpatents

    Forsberg, C.W.

    1998-11-03

    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.

  9. Depleted uranium as a backfill for nuclear fuel waste package

    SciTech Connect

    Forsberg, Charles W.

    1997-12-01

    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 isotonically-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.

  10. A cermet fuel reactor for nuclear thermal propulsion

    NASA Technical Reports Server (NTRS)

    Kruger, Gordon

    1991-01-01

    Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

  11. Cosmic ray muons for spent nuclear fuel monitoring

    NASA Astrophysics Data System (ADS)

    Chatzidakis, Stylianos

    There is a steady increase in the volume of spent nuclear fuel stored on-site (at reactor) as currently there is no permanent disposal option. No alternative disposal path is available and storage of spent nuclear fuel in dry storage containers is anticipated for the near future. In this dissertation, a capability to monitor spent nuclear fuel stored within dry casks using cosmic ray muons is developed. The motivation stems from the need to investigate whether the stored content agrees with facility declarations to allow proliferation detection and international treaty verification. Cosmic ray muons are charged particles generated naturally in the atmosphere from high energy cosmic rays. Using muons for proliferation detection and international treaty verification of spent nuclear fuel is a novel approach to nuclear security that presents significant advantages. Among others, muons have the ability to penetrate high density materials, are freely available, no radiological sources are required and consequently there is a total absence of any artificial radiological dose. A methodology is developed to demonstrate the applicability of muons for nuclear nonproliferation monitoring of spent nuclear fuel dry casks. Purpose is to use muons to differentiate between spent nuclear fuel dry casks with different amount of loading, not feasible with any other technique. Muon scattering and transmission are used to perform monitoring and imaging of the stored contents of dry casks loaded with spent nuclear fuel. It is shown that one missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the scattering distributions with 300,000 muons or more. A Bayesian monitoring algorithm was derived to allow differentiation of a fully loaded dry cask from one with a fuel assembly missing in the order of minutes and negligible error rate. Muon scattering and transmission simulations are used to reconstruct the stored contents of sealed dry casks

  12. A method for monitoring nuclear absorption coefficients of aviation fuels

    NASA Technical Reports Server (NTRS)

    Sprinkle, Danny R.; Shen, Chih-Ping

    1989-01-01

    A technique for monitoring variability in the nuclear absorption characteristics of aviation fuels has been developed. It is based on a highly collimated low energy gamma radiation source and a sodium iodide counter. The source and the counter assembly are separated by a geometrically well-defined test fuel cell. A computer program for determining the mass attenuation coefficient of the test fuel sample, based on the data acquired for a preset counting period, has been developed and tested on several types of aviation fuel.

  13. Nuclear Fuel Leasing, Recycling and proliferation: Modeling a Global View

    SciTech Connect

    Crozat, M P; Choi, J; Reis, V H; Hill, R

    2004-03-10

    On February 11, 2004, U.S. President George W. Bush, in a speech to the National Defense University stated: ''The world must create a safe, orderly system to field civilian nuclear plants without adding to the danger of weapons proliferation. The world's leading nuclear exporters should ensure that states have reliable access at reasonable cost to fuel for civilian reactors, so long as those states renounce enrichment and reprocessing. Enrichment and reprocessing are not necessary for nations seeking to harness nuclear energy for peaceful purposes.'' This concept would require nations to choose one of two paths for civilian nuclear development: those that only have reactors and those that contain one or more elements of the nuclear fuel cycle, including recycling. ''Fuel cycle'' states would enrich uranium, manufacture and lease fuel to ''reactor'' states and receive the reactor states' spent fuel. All parties would accede to stringent security and safeguard standards, embedded within a newly invigorated international regime. Reactor states would be relieved of the financial, environmental (and political) burden of enriching and manufacturing fuel and dealing with spent fuel. Fuel cycle states would potentially earn money on leasing the fuel and perhaps on sales of reactors to the reactor states. Such a leasing concept is especially interesting in scenarios which envision growth in nuclear power, and an important consideration for such a nuclear growth regime is the role of recycling of civilian spent fuel. Recycling holds promise for improved management of spent fuel and efficient utilization of resources, but continues to raise the specter of a world with uncontrolled nuclear weapons proliferation. If done effectively, a fuel-leasing concept could help create a political and economic foundation for significant growth of clean, carbon-free nuclear power while providing a mechanism for significant international cooperation to reduce proliferation concern. This

  14. Breeding nuclear fuels with accelerators: replacement for breeder reactors

    SciTech Connect

    Grand, P.; Takahashi, H.

    1984-01-01

    One application of high energy particle accelerators has been, and still is, the production of nuclear fuel for the nuclear energy industry; tantalizing because it would create a whole new industry. This approach to producing fissile from fertile material was first considered in the early 1950's in the context of the nuclear weapons program. A considerable development effort was expended before discovery of uranium ore in New Mexico put an end to the project. Later, US commitment to the Liquid Metal Fast Breeder Reactors (LMFBR) killed any further interest in pursuing accelerator breeder technology. Interest in the application of accelerators to breed nuclear fuels, and possibly burn nuclear wastes, revived in the late 1970's, when the LMFBR came under attack during the Carter administration. This period gave the opportunity to revisit the concept in view of the present state of the technology. This evaluation and the extensive calculational modeling of target designs that have been carried out are promising. In fact, a nuclear fuel cycle of Light Water Reactors and Accelerator Breeders is competitive to that of the LMFBR. At this time, however, the relative abundance of uranium reserves vs electricity demand and projected growth rate render this study purely academic. It will be for the next generation of accelerator builders to demonstate the competitiveness of this technology versus that of other nuclear fuel cycles, such as LMFBR's or Fusion Hybrid systems. 22 references, 1 figure, 5 tables.

  15. Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels

    SciTech Connect

    Todd R. Allen, Director

    2011-04-01

    The Office of Science, Basic Energy Sciences, has funded the INL as one of the Energy Frontier Research Centers in the area of material science of nuclear fuels. This document is the required annual report to the Office of Science that outlines the accomplishments for the period of May 2010 through April 2011. The aim of the Center for Material Science of Nuclear Fuels (CMSNF) is to establish the foundation for predictive understanding of the effects of irradiation-induced defects on thermal transport in oxide nuclear fuels. The science driver of the center’s investigation is to understand how complex defect and microstructures affect phonon mediated thermal transport in UO2, and achieve this understanding for the particular case of irradiation-induced defects and microstructures. The center’s research thus includes modeling and measurement of thermal transport in oxide fuels with different levels of impurities, lattice disorder and irradiation-induced microstructure, as well as theoretical and experimental investigation of the evolution of disorder, stoichiometry and microstructure in nuclear fuel under irradiation. With the premise that thermal transport in irradiated UO2 is a phonon-mediated energy transport process in a crystalline material with defects and microstructure, a step-by-step approach will be utilized to understand the effects of types of defects and microstructures on the collective phonon dynamics in irradiated UO2. Our efforts under the thermal transport thrust involved both measurement of diffusive phonon transport (an approach that integrates over the entire phonon spectrum) and spectroscopic measurements of phonon attenuation/lifetime and phonon dispersion. Our distinct experimental efforts dovetail with our modeling effort involving atomistic simulation of phonon transport and prediction of lattice thermal conductivity using the Boltzmann transport framework.

  16. An analysis of international nuclear fuel supply options

    NASA Astrophysics Data System (ADS)

    Taylor, J'tia Patrice

    As the global demand for energy grows, many nations are considering developing or increasing nuclear capacity as a viable, long-term power source. To assess the possible expansion of nuclear power and the intricate relationships---which cover the range of economics, security, and material supply and demand---between established and aspirant nuclear generating entities requires models and system analysis tools that integrate all aspects of the nuclear enterprise. Computational tools and methods now exist across diverse research areas, such as operations research and nuclear engineering, to develop such a tool. This dissertation aims to develop methodologies and employ and expand on existing sources to develop a multipurpose tool to analyze international nuclear fuel supply options. The dissertation is comprised of two distinct components: the development of the Material, Economics, and Proliferation Assessment Tool (MEPAT), and analysis of fuel cycle scenarios using the tool. Development of MEPAT is aimed for unrestricted distribution and therefore uses publicly available and open-source codes in its development when possible. MEPAT is built using the Powersim Studio platform that is widely used in systems analysis. MEPAT development is divided into three modules focusing on: material movement; nonproliferation; and economics. The material movement module tracks material quantity in each process of the fuel cycle and in each nuclear program with respect to ownership, location and composition. The material movement module builds on techniques employed by fuel cycle models such as the Verifiable Fuel Cycle Simulation (VISION) code developed at the Idaho National Laboratory under the Advanced Fuel Cycle Initiative (AFCI) for the analysis of domestic fuel cycle. Material movement parameters such as lending and reactor preference, as well as fuel cycle parameters such as process times and material factors are user-specified through a Microsoft Excel(c) data spreadsheet

  17. Fleet Conversion in Local Government: Determinants of Driver Fuel Choice for Bi-Fuel Vehicles

    ERIC Educational Resources Information Center

    Johns, Kimberly D.; Khovanova, Kseniya M.; Welch, Eric W.

    2009-01-01

    This study evaluates the conversion of one local government's fleet from gasoline to bi-fuel E-85, compressed natural gas, and liquid propane gas powered vehicles at the midpoint of a 10-year conversion plan. This study employs a behavioral model based on the theory of reasoned action to explore factors that influence an individual's perceived and…

  18. An evaluation of the nuclear fuel performance code BISON

    SciTech Connect

    Perez, D. M.; Williamson, R. L.; Novascone, S. R.; Larson, T. K.; Hales, J. D.; Spencer, B. W.; Pastore, G.

    2013-07-01

    BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. The code is applicable to both steady and transient fuel behavior and is used to analyze either 2D axisymmetric or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods and other well known fuel performance codes. Results from several assessment cases are reported, with emphasis on fuel centerline temperatures at various stages of fuel life, fission gas release, and clad deformation during pellet clad mechanical interaction (PCMI). BISON comparisons to fuel centerline temperature measurements are very good at beginning of life and reasonable at high burnup. Although limited to date, fission gas release comparisons are very good. Comparisons of rod diameter following significant power ramping are also good and demonstrate BISON's unique ability to model discrete pellet behavior and accurately predict clad ridging from PCMI. (authors)

  19. Separation of actinides from spent nuclear fuel: A review.

    PubMed

    Veliscek-Carolan, Jessica

    2016-11-15

    This review summarises the methods currently available to extract radioactive actinide elements from solutions of spent nuclear fuel. This separation of actinides reduces the hazards associated with spent nuclear fuel, such as its radiotoxicity, volume and the amount of time required for its' radioactivity to return to naturally occurring levels. Separation of actinides from environmental water systems is also briefly discussed. The actinide elements typically found in spent nuclear fuel include uranium, plutonium and the minor actinides (americium, neptunium and curium). Separation methods for uranium and plutonium are reasonably well established. On the other hand separation of the minor actinides from lanthanide fission products also present in spent nuclear fuel is an ongoing challenge and an area of active research. Several separation methods for selective removal of these actinides from spent nuclear fuel will be described. These separation methods include solvent extraction, which is the most commonly used method for radiochemical separations, as well as the less developed but promising use of adsorption and ion-exchange materials.

  20. A framework and methodology for nuclear fuel cycle transparency.

    SciTech Connect

    McClellan, Yvonne; York, David L.; Inoue, Naoko; Love, Tracia L.; Rochau, Gary Eugene

    2006-02-01

    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.

  1. Ablation and deceleration of mass-driver launched projectiles for space disposal of nuclear wastes

    NASA Technical Reports Server (NTRS)

    Park, C.; Bowen, S. W.

    1981-01-01

    The energy cost of launching a projectile containing nuclear waste is two orders of magnitude lower with a mass driver than with a typical rocket system. A mass driver scheme will be feasible, however, only if ablation and deceleration are within certain tolerable limits. It is shown that if a hemisphere-cylinder-shaped projectile protected thermally with a graphite nose is launched vertically to attain a velocity of 17 km/sec at an altitude of 40 km, the mass loss from ablation during atmospheric flight will be less than 0.1 ton, provided the radius of the projectile is under 20 cm and the projectile's mass is of the order of 1 ton. The velocity loss from drag will vary from 0.4 to 30 km/sec, depending on the mass and radius of the projectile, the smaller velocity loss corresponding to large mass and small radius. Ablation is always within a tolerable range for schemes using a mass driver launcher to dispose of nuclear wastes outside the solar system. Deceleration can also be held in the tolerable range if the mass and diameter of the projectile are properly chosen.

  2. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... advance notification of shipment of irradiated reactor fuel or nuclear waste must contain the following... irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel...

  3. Comparative Analysis on Nuclear Fuel Sustainability Aspect of FBR

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Irwanto, Dwi; Suzuki, Mitsutoshi; Saito, Masaki

    2017-01-01

    Recycle program of spent nuclear fuel (SNF) will have some challanges in term of fuel cycle capability and its facilities as well as nuclear non-proliferation concern of special nuclear materials. A different analysis approach as a comparative study have been analyzed based on breeding ratio and heavy metal inventory ratio concepts in fast breeder reactor (FBR) type. Breeding ratio and heavy metal inventory obtain higher than unity which shows breeding gain or surplus inventory of heavy metals are obtained. Breeding ratio indicates the fuel conversion capability from conversion process of fertile materials into fissile material such as fertile materials of U-238, Pu-238, Pu-240 and fissile materials of Pu-239 and Pu-241. Inventory ratio approaches are appropriate to estimate some selected actinide as a mass inventory production such as plutonium inventory ratio which estimate the surplus mass inventory from the ratio of produced plutonium at the net of operation to the initial inventory ratio.

  4. International nuclear fuel cycle fact book: Revision 9

    SciTech Connect

    Leigh, I.W.

    1989-01-01

    The International Nuclear Fuel Cycle Fact Book has been compiled in an effort to provide current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. The Fact Book contains: national summaries in which a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; and international agencies in which a section for each of the international agencies which has significant fuel cycle involvement, and a listing of nuclear societies. The national summaries, in addition to the data described above, feature a small map for each country as well as some general information. The latter is presented from the perspective of the Fact Book user in the United States.

  5. International nuclear fuel cycle fact book. [Contains glossary

    SciTech Connect

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

    1987-01-01

    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.

  6. International Nuclear Fuel Cycle Fact Book. Revision 12

    SciTech Connect

    Leigh, I.W.

    1992-05-01

    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.

  7. Secondary ion mass spectrometry of irradiated nuclear fuel and cladding

    NASA Astrophysics Data System (ADS)

    Portier, S.; Brémier, S.; Walker, C. T.

    2007-06-01

    The principles and operating modes of secondary ion mass spectrometry (SIMS) are first described after which the different methods of quantification are summarised. Some current applications of SIMS in nuclear fuel and cladding research are then reviewed after briefly considering the modifications that are needed to allow a SIMS instrument to be used for the analysis of highly radioactive materials. Amongst the applications reported are the investigation of the behaviour of fission gas xenon and the volatile fission products tellurium, iodine and caesium in UO2 nuclear fuel, measurement of the radial distribution of Pu isotopes in mixed oxide (MOX) fuel and of the radial distribution of Gd isotopes in (U,Gd)O2 fuel, and determination of the distribution of Li and B in the external oxide layer on Zircaloy cladding. It is evident from the large amount of new information gained that SIMS is a powerful complementary technique to electron probe microanalysis (EPMA) in these fields of study.

  8. Temperature measuring analysis of the nuclear reactor fuel assembly

    SciTech Connect

    Urban, F. E-mail: zdenko.zavodny@stuba.sk; Kučák, L. E-mail: zdenko.zavodny@stuba.sk; Bereznai, J. E-mail: zdenko.zavodny@stuba.sk; Závodný, Z. E-mail: zdenko.zavodny@stuba.sk; Muškát, P. E-mail: zdenko.zavodny@stuba.sk

    2014-08-06

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  9. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in...

  10. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in...

  11. Modeling Deep Burn TRISO particle nuclear fuel

    NASA Astrophysics Data System (ADS)

    Besmann, T. M.; Stoller, R. E.; Samolyuk, G.; Schuck, P. C.; Golubov, S. I.; Rudin, S. P.; Wills, J. M.; Coe, J. D.; Wirth, B. D.; Kim, S.; Morgan, D. D.; Szlufarska, I.

    2012-11-01

    Under the DOE Deep Burn program TRISO fuel is being investigated as a fuel form for consuming plutonium and minor actinides, and for greater efficiency in uranium utilization. The result will thus be to drive TRISO particulate fuel to very high burn-ups. In the current effort the various phenomena in the TRISO particle are being modeled using a variety of techniques. The chemical behavior is being treated utilizing thermochemical analysis to identify phase formation/transformation and chemical activities in the particle, including kernel migration. Density functional theory is being used to understand fission product diffusion within the plutonia oxide kernel, the fission product's attack on the SiC coating layer, as well as fission product diffusion through an alternative coating layer, ZrC. Finally, a multiscale approach is being used to understand thermal transport, including the effect of radiation damage induced defects, in a model SiC material.

  12. Double-clad nuclear-fuel safety rod

    DOEpatents

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

    1981-12-30

    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.

  13. Apparatus for injection casting metallic nuclear energy fuel rods

    DOEpatents

    Seidel, Bobby R.; Tracy, Donald B.; Griffiths, Vernon

    1991-01-01

    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.

  14. Spent nuclear fuel Canister Storage Building CDR Review Committee report

    SciTech Connect

    Dana, W.P.

    1995-12-01

    The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

  15. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOEpatents

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  16. Synergistic smart fuel for in-pile nuclear reactor measurements

    SciTech Connect

    Smith, J.A.; Kotter, D.K.; Ali, R.A.; Garrett, S.L.

    2013-07-01

    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.

  17. Signatures of Extended Storage of Used Nuclear Fuel in Casks

    SciTech Connect

    Rauch, Eric Benton

    2016-09-28

    As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Control Technologies (MPACT) campaign. Under this project we looked at fingerprinting each cask's neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.

  18. Microbial Biofilm Growth on Irradiated, Spent Nuclear Fuel Cladding

    SciTech Connect

    S.M. Frank

    2009-02-01

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 × 103 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments.

  19. Preparation of nuclear fuel spheres by flotation-internal gelation

    DOEpatents

    Haas, Paul A.; Fowler, Victor L.; Lloyd, Milton H.

    1987-01-01

    A simplified internal gelation process for the preparation of gel spheres of nuclear fuels. The process utilizes perchloroethylene as a gelation medium. Gelation is accomplished by directing droplets of a nuclear fuel broth into a moving volume of hot perchloroethylene (about 85.degree. C.) in a trough. Gelation takes place as the droplets float on the surface of the perchloroethylene and the resultant gel spheres are carried directly into an ager column which is attached to the trough. The aged spheres are disengaged from the perchloroethylene on a moving screen and are deposited in an aqueous wash column.

  20. Preparation of nuclear fuel spheres by flotation-internal gelation

    DOEpatents

    Haas, P.A.; Fowler, V.L.; Lloyd, M.H.

    1984-12-21

    A simplified internal gelation process is claimed for the preparation of gel spheres of nuclear fuels. The process utilizes perchloroethylene as a gelation medium. Gelation is accomplished by directing droplets of a nuclear fuel broth into a moving volume of hot perchloroethylene (about 85/sup 0/C) in a trough. Gelation takes place as the droplets float on the surface of the perchloroethylene and the resultant gel spheres are carried directly into an ager column which is attached to the trough. The aged spheres are disengaged from the perchloroethylene on a moving screen and are deposited in an aqueous wash column. 3 figs.

  1. Cost and availability of gadolinium for nuclear fuel reprocessing plants

    SciTech Connect

    Klepper, O.H.

    1985-06-01

    Gadolinium is currently planned for use as a soluble neutron poison in nuclear fuel reprocessing plants to prevent criticality of solutions of spent fuel. Gadolinium is relatively rare and expensive. The present study was undertaken therefore to estimate whether this material is likely to be available in quantities sufficient for fuel reprocessing and at reasonable prices. It was found that gadolinium, one of 16 rare earth elements, appears in the marketplace as a by-product and that its present supply is a function of the production rate of other more prevalent rare earths. The potential demand for gadolinium in a fuel reprocessing facility serving a future fast reactor industry amounts to only a small fraction of the supply. At the present rate of consumption, domestic supplies of rare earths containing gadolinium are adequate to meet national needs (including fuel reprocessing) for over 100 years. With access to foreign sources, US demands can be met well beyond the 21st century. It is concluded therefore that the supply of gadolinium will quite likely be more than adequate for reprocessing spent fuel for the early generation of fast reactors. The current price of 99.99% pure gadolinium oxide lies in the range $50/lb to $65/lb (1984 dollars). By the year 2020, in time for reprocessing spent fuel from an early generation of large fast reactors, the corresponding values are expected to lie in the $60/lb to $75/lb (1984 dollars) price range. This increase is modest and its economic impact on nuclear fuel reprocessing would be minor. The economic potential for recovering gadolinium from the wastes of nuclear fuel reprocessing plants (which use gadolinium neutron poison) was also investigated. The cost of recycled gadolinium was estimated at over twelve times the cost of fresh gadolinium, and thus recycle using current recovery technology is not economical. 15 refs., 4 figs., 11 tabs.

  2. Technology Insights and Perspectives for Nuclear Fuel Cycle Concepts

    SciTech Connect

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

    2010-09-01

    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

  3. Quantitative assessment of proposals on assurance of nuclear fuel supply

    SciTech Connect

    Tanaka, T.; Kuno, Y.; Tanaka, S.

    2013-07-01

    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.

  4. Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel

    SciTech Connect

    Marshall, William BJ J; Wagner, John C

    2012-01-01

    Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., <40 years). Extended storage (ES) time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k{sub eff}. Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

  5. Method for storing nuclear fuel in respositories

    DOEpatents

    Schweitzer, D.G.; Sastre, C.

    A method for storing radioactive spent fuel in repositories containing polyphenyl or silicon oil as the storage medium is disclosed. Polyphenyls and silicon oils are non-corrosive and are not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

  6. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    SciTech Connect

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A. Ignatiev, V. V.; Subbotin, S. A. Tsibulskiy, V. F.

    2015-12-15

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  7. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    NASA Astrophysics Data System (ADS)

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-12-01

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  8. Manufacture of bonded-particle nuclear fuel composites

    DOEpatents

    Stradley, J.G.; Sease, J.D.

    1973-10-01

    A preselected volume of nuclear fuel particles are placed in a cylindrical mold cavity followed by a solid pellet of resin--carbon matrix material of preselected volume. The mold is heated to liquefy the pellet and the liquefied matrix forced throughout the interstices of the fuel particles by advancing a piston into the mold cavity. Excess matrix is permitted to escape through a vent hole in the end of the mold opposite to that end where the pellet was originally disposed. After the matrix is resolidified by cooling, the resultant fuel composite is removed from the mold and the resin component of the matrix carbonized. (Official Gazette)

  9. Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications

    SciTech Connect

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

    2011-11-01

    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.

  10. Nuclear Fuel Cycle Reasoner: PNNL FY13 Report

    SciTech Connect

    Hohimer, Ryan E.; Strasburg, Jana D.

    2013-09-30

    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.

  11. Dissolution of Spent Nuclear Fuel in Carbonate-Peroxide Solution

    SciTech Connect

    Soderquist, Chuck Z.; Hanson, Brady D.

    2010-01-31

    This study shows that spent UO2 fuel can be completely dissolved in a carbonate-peroxide solution apparently without attacking the metallic Mo-Tc-Ru-Rh-Pd fission product phase. Samples of spent nuclear fuel were pulverized and sieved to a uniform size, then duplicate aliquots were weighed into beakers for analysis. One set was dissolved in near-boiling 10M nitric acid, and the other set was dissolved in a solution of ammonium carbonate and hydrogen peroxide at room temperature. All the resulting fuel solutions were then analyzed for Sr-90, Tc-99, Cs-137, plutonium, and Am-241. For all the samples, the concentrations of Cs-137, Sr-90, plutonium, and Am-241 were the same for both the nitric acid dissolution and the ammonium carbonate-hydrogen peroxide dissolution, but the technetium concentration of the ammonium carbonate-hydrogen peroxide fuel solution was only about 25% of the same fuels dissolved in hot nitric acid.

  12. Thermo-Elastic Finite Element Analyses of Annular Nuclear Fuels

    NASA Astrophysics Data System (ADS)

    Kwon, Y. D.; Kwon, S. B.; Rho, K. T.; Kim, M. S.; Song, H. J.

    In this study, we tried to examine the pros and cons of the annular type of fuel concerning mainly with the temperatures and stresses of pellet and cladding. The inner and outer gaps between pellet and cladding may play an important role on the temperature distribution and stress distribution of fuel system. Thus, we tested several inner and outer gap cases, and we evaluated the effect of gaps on fuel systems. We conducted thermo-elastic-plastic-creep analyses using an in-house thermo-elastic-plastic-creep finite element program that adopted the 'effective-stress-function' algorithm. Most analyses were conducted until the gaps disappeared; however, certain analyses lasted for 1582 days, after which the fuels were replaced. Further study on the optimal gaps sizes for annular nuclear fuel systems is still required.

  13. Characterization of Nuclear Fuel using Multivariate Statistical Analysis

    SciTech Connect

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

    2007-11-27

    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.

  14. Nuclear fuel elements and method of making same

    DOEpatents

    Schweitzer, Donald G.

    1992-01-01

    A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

  15. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    SciTech Connect

    Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs

    2014-11-01

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

  16. Sensitivity analysis and optimization of the nuclear fuel cycle

    SciTech Connect

    Passerini, S.; Kazimi, M. S.; Shwageraus, E.

    2012-07-01

    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)

  17. Zone approaches to international safeguards of a nuclear fuel cycle

    SciTech Connect

    Fishbone, L.G.; Higinbotham, W.A.

    1986-01-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the results of safeguards verifications for the individual facilities within it. We have examined safeguards approaches for a state nuclear fuel cycle that take into account the existence of all of the nuclear facilities in the state. We have focussed on the fresh-fuel zone of an advanced nuclear fuel cycle, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. The intention is to develop an approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the zone approach and for some reasonable intermediate safeguards approaches. Technical effectiveness, in these cases, means an estimate of the assurance that all nuclear material has been accounted for.

  18. Modelling oxidation behaviour in operating defective nuclear reactor fuel elements

    NASA Astrophysics Data System (ADS)

    Higgs, Jamie D.

    CANDU nuclear reactors are powered by ceramic uranium dioxide (UO 2) fuel pellets encased in a zirconium-alloy sheath. Occasionally, holes develop in the sheath, allowing steam ingress into the fuel-to-sheath gap, thus exposing the fuel to an oxidizing environment. Oxidation of UO2 fuel may lead to a reduction of fuel thermal conductivity and melting point, both reducing the margin to prevent fuel centre-line melting during transient or even normal operating conditions. Along with increasing fuel temperature, fuel oxidation also enhances the release of radioactive fission products into the reactor coolant. For the first time, a mechanistic treatment has been considered to predict fuel oxidation behaviour in operating defective fuel elements by coupling fuel oxidation kinetics, interstitial oxygen diffusion and heat transfer with sheath oxidation and hydriding rates and gas phase transport in both the fuel-to-sheath gap and within the fuel cracks. The three highly non-linear phenomena (solid-state oxygen diffusion, gas-phase transport and heat transfer) coupled in this treatment were modelled using a finite element technique. The result is a numerical tool that can provide predictions of both the temperature and oxygen-to-uranium (O/U) ratio profile both radially and axially along the fuel element length. The two-dimensional (azimuthally-symmetric) model has been compared to oxygen profile measurements from commercial reactor defective fuel with operating linear power ratings ranging from 26 to 51 kW m-1. Model predictions agree well with experimental observations. Defect size, linear power rating and post-defect residence time (PDRT) appear to be the factors that most influence the extent and rate of fuel oxidation. Thermodynamic modelling of hyperstoichiometric fuel provided the boundary conditions for the fuel oxidation kinetics model. A refined thermodynamic treatment for hyperstoichiometric UO2 has been established. Neutron diffraction experiments at Los Alamos

  19. Air Shipment of Spent Nuclear Fuel from Romania to Russia

    SciTech Connect

    Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

    2010-10-01

    Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

  20. Systems for the Intermodal Routing of Spent Nuclear Fuel

    SciTech Connect

    Peterson, Steven K; Liu, Cheng

    2015-01-01

    The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable system for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of

  1. Nuclear design of Helical Cruciform Fuel rods

    SciTech Connect

    Shirvan, K.; Kazimi, M. S.

    2012-07-01

    In order to increase the power density of current and new light water reactor designs, the Helical Cruciform Fuel (HCF) rods are proposed. The HCF rods are equivalent to a cylindrical rod, with the fuel in a cruciform shaped, twisted axially. The HCF rods increase the surface area to volume ratio and inter-subchannel mixing behavior due to their cruciform and helical shapes, respectively. In a previous study, the HCF rods have shown the potential to up-rate existing PWRs by 50% and BWRs by 25%. However, HCF rods do display different neutronics modeling and performance. The cruciform cross section of HCF rods creates radially asymmetric heat generation and temperature distribution. The nominal HCF rod's beginning of life reactivity is reduced, compared to a cylindrical rod with the same fuel volume, by 500 pcm, due to increase in absorption in cladding. The rotation of these rods accounts for reactivity changes, which depends on the H/HM ratio of the pin cell. The HCF geometry shows large sensitivities to U{sup 235} or gadolinium enrichments compared to a cylindrical geometry. In addition, the gadolinium-containing HCF rods show a stronger effect on neighboring HCF rods than in case of cylindrical rods, depending on the orientation of the HCF rods. The helical geometry of the rods introduces axial shadowing of about 600 pcm, not seen in typical cylindrical rods. (authors)

  2. Spallator and APEX nuclear fuel cycle: a new option for nuclear power

    SciTech Connect

    Steinberg, M.

    1982-01-01

    A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high-energy (1 to 2 GeV) protons on a heavy-metal target. The neutrons are absorbed in a surrounding natural-uranium or thorium blanket in which fissile Pu-239 to U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high-beam-current continuous-wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a self-sufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of short-lived fission products external to the fuel cycle eliminates the need for long-term geological age shortage of fission-product waste.

  3. Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

    SciTech Connect

    Weber, William J.; Zhang, Yanwen

    2016-09-20

    This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effects of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied.

  4. The growing problem of stranded used nuclear fuel.

    PubMed

    Alley, William M; Alley, Rosemarie

    2014-02-18

    By 2050, almost all U.S. nuclear reactors will have reached their 60 year maximum expected life. Many will shut down sooner. With no assurance that the current approach for finding a geologic repository or interim storage sites will succeed, used nuclear fuel could be stranded indefinitely at more than 70 sites in 35 states. Societal discussions about the future of nuclear waste should be framed in terms of the relative risks of all alternatives. We review and compare onsite storage, interim storage, and a geologic repository, as well as how these alternatives are presented to the public.

  5. Nuclear Fuel Cycle Reasoner: PNNL FY12 Report

    SciTech Connect

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

    2013-05-03

    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.

  6. Safety analysis of irradiated nuclear fuel transportation container

    SciTech Connect

    Uspuras, E.; Rimkevicius, S.

    2007-07-01

    Ignalina NPP comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The Safety Analysis Report (SAR) was developed to demonstrate that the proposed set of equipment performs all functions and assures the required level of safety for both normal operation and accident conditions. The purpose of this paper is to introduce the content and main results of SAR, focusing attention on the container used to transport spent fuel assemblies from Unit I on Unit 2. In the SAR, the structural integrity, thermal, radiological and nuclear safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel transportation container and other equipment are in compliance with functional, design and regulatory requirements and assure the required safety level. (authors)

  7. METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Layer, E.H. Jr.; Peet, C.S.

    1962-01-23

    A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

  8. Method of controlling crystallite size in nuclear-reactor fuels

    DOEpatents

    Lloyd, Milton H.; Collins, Jack L.; Shell, Sam E.

    1985-01-01

    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.

  9. Spent nuclear fuel project design basis capacity study

    SciTech Connect

    Cleveland, K.J.

    1996-09-09

    A parametric study of the Spent Nuclear Fuel Project system capacity is presented. The study was completed using a commercially available software package to develop a summary level model of the major project systems. Alternative configurations, sub-system cycle times, and operating scenarios were tested to identify their impact on total project duration and equipment requirements.

  10. Methods and apparatuses for the development of microstructured nuclear fuels

    DOEpatents

    Jarvinen, Gordon D.; Carroll, David W.; Devlin, David J.

    2009-04-21

    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.

  11. Method of controlling crystallite size in nuclear-reactor fuels

    DOEpatents

    Lloyd, M.H.; Collins, J.L.; Shell, S.E.

    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.

  12. MOX LTA Fuel Cycle Analyses: Nuclear and Radiation Safety

    SciTech Connect

    Pavlovitchev, A.M.

    2001-09-28

    Tasks of nuclear safety assurance for storage and transport of fresh mixed uranium-plutonium fuel of the VVER-1000 reactor are considered in the view of 3 MOX LTAs introduction into the core. The precise code MCU that realizes the Monte Carlo method is used for calculations.

  13. METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Hauth, J.J.; Anicetti, R.J.

    1962-12-01

    A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

  14. DUCTILE URANIUM FUEL FOR NUCLEAR REACTORS AND METHOD OF MAKING

    DOEpatents

    Zegler, S.T.

    1963-11-01

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

  15. Review of Drying Methods for Spent Nuclear Fuel

    SciTech Connect

    Large, W.S.

    1999-10-21

    SRTC is developing technology for direct disposal of aluminum spent nuclear fuel (SNF). The development program includes analyses and tests to support design and safe operation of a facility for ''road ready'' dry storage of SNF-filled canisters. The current technology development plan includes review of available SNF drying methods and recommendation of a drying method for aluminum SNF.

  16. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2010-03-05

    with uranium to make mixed-oxide ( MOX ) fuel, in which the 239Pu largely substitutes for 235U. Two French reprocessing plants at La Hague can each...and France also have older plants to reprocess gas-cooled reactor fuel, and India has a 275-ton plant.53 About 200 metric tons of MOX fuel is used...to make MOX fuel for today’s nuclear power plants are modest. Existing commercial light water reactors use ordinary water to slow down, or “moderate

  17. Non-proliferation aspects of commercial nuclear fuel cycles

    SciTech Connect

    Persiani, P.J.

    1992-03-01

    In the international non-proliferation regime of concern, the denatured uranium fuel cycles should be assessed in terms of proliferation criteria. The study attempts to establish a technical basis for a comparative assessment of the proliferation potential among the denatured uranium, the plutonium-uranium, and the breeder fuel cycles. The international commercial fuel cycles were analyzed within the context of current technology for the potential to: divert and upgrade the quality of nuclear material via the construction and operation of clandestine-scale reprocessing facilities and low-technology electromagnetic enrichment facilities, such as calutrons. Eleven fuel cycles were analyzed to determine the number of calutron base-feed units (kg of heavy metal) contained in typical fuel assemblies for each of the cycles to produce a significant quantity of high enriched uranium (HEU) material within limited time periods. The study indicates that, for calutron specifications within current technology, the level of effort to enrich 3% {sup 235}U/{sup 238}U fuel is a factor of 20 lower than the two-stage process with natural uranium feed; 16 to 20 calutrons and 6 to 8 fuel assemblies are required to produce a significant quantity of HEU in one year of operation. Consequently, non-proliferation concerns should be emphasized at the front-end of the fuel cycles involving uranium as well as at the back-end involving plutonium.

  18. The role of accelerators in the nuclear fuel cycle

    SciTech Connect

    Takahashi, Hiroshi.

    1990-01-01

    The use of neutrons produced by the medium energy proton accelerator (1 GeV--3 GeV) has considerable potential in reconstructing the nuclear fuel cycle. About 1.5 {approximately} 2.5 ton of fissile material can be produced annually by injecting a 450 MW proton beam directly into fertile materials. A source of neutrons, produced by a proton beam, to supply subcritical reactors could alleviate many of the safety problems associated with critical assemblies, such as positive reactivity coefficients due to coolant voiding. The transient power of the target can be swiftly controlled by controlling the power of the proton beam. Also, the use of a proton beam would allow more flexibility in the choice of fuel and structural materials which otherwise might reduce the reactivity of reactors. This paper discusses the rate of accelerators in the transmutation of radioactive wastes of the nuclear fuel cycles. 34 refs., 17 figs., 9 tabs.

  19. Heat Transfer Modeling of Dry Spent Nuclear Fuel Storage Facilities

    SciTech Connect

    Lee, S.Y.

    1999-01-13

    The present work was undertaken to provide heat transfer model that accurately predicts the thermal performance of dry spent nuclear fuel storage facilities. One of the storage configurations being considered for DOE Aluminum-clad Spent Nuclear Fuel (Al-SNF), such as the Material and Testing Reactor (MTR) fuel, is in a dry storage facility. To support design studies of storage options a computational and experimental program has been conducted at the Savannah River Site (SRS). The main objective is to develop heat transfer models including natural convection effects internal to an interim dry storage canister and to geological codisposal Waste Package (WP). Calculated temperatures will be used to demonstrate engineering viability of a dry storage option in enclosed interim storage and geological repository WP and to assess the chemical and physical behaviors of the Al-SNF in the dry storage facilities. The current paper describes the modeling approaches and presents the computational results along with the experimental data.

  20. Modeling and Simulation of a Nuclear Fuel Element Test Section

    NASA Technical Reports Server (NTRS)

    Moran, Robert P.; Emrich, William

    2011-01-01

    "The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

  1. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    SciTech Connect

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01

    Fuel handling development testing 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. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems` Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment.

  2. Enduring Nuclear Fuel Cycle, Proceedings of a panel discussion

    SciTech Connect

    Walter, C. E., LLNL

    1997-11-18

    The panel reviewed the complete nuclear fuel cycle in the context of alternate energy resources, energy need projections, effects on the environment, susceptibility of nuclear materials to theft, diversion, and weapon proliferation. We also looked at ethical considerations of energy use, as well as waste, and its effects. The scope of the review extended to the end of the next century with due regard for world populations beyond that period. The intent was to take a long- range view and to project, not forecast, the future based on ethical rationales, and to avoid, as often happens, long-range discussions that quickly zoom in on only the next few decades. A specific nuclear fuel cycle technology that could satisfy these considerations was described and can be applied globally.

  3. Interim report spent nuclear fuel retrieval system fuel handling development testing

    SciTech Connect

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

    1997-06-01

    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.

  4. Evaluation of conventional power systems. [emphasizing fossil fuels and nuclear energy

    NASA Technical Reports Server (NTRS)

    Smith, K. R.; Weyant, J.; Holdren, J. P.

    1975-01-01

    The technical, economic, and environmental characteristics of (thermal, nonsolar) electric power plants are reviewed. The fuel cycle, from extraction of new fuel to final waste management, is included. Emphasis is placed on the fossil fuel and nuclear technologies.

  5. Laser Shockwave Technique For Characterization Of Nuclear Fuel Plate Interfaces

    SciTech Connect

    James A. Smith; Barry H. Rabin; Mathieu Perton; Daniel Lévesque; Jean-Pierre Monchalin; Martin Lord

    2012-07-01

    The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

  6. Retrievable fuel pin end member for a nuclear reactor

    DOEpatents

    Rosa, Jerry M.

    1982-01-01

    A bottom end member (17b) on a retrievable fuel pin (13b) secures the pin (13b) within a nuclear reactor (12) by engaging on a transverse attachment rail (18) with a spring clip type of action. Removal and reinstallation if facilitated as only axial movement of the fuel pin (13b) is required for either operation. A pair of resilient axially extending blades (31) are spaced apart to define a slot (24) having a seat region (34) which receives the rail (18) and having a land region (37), closer to the tips (39) of the blades (31) which is normally of less width than the rail (18). Thus an axially directed force sufficient to wedge the resilient blades (31) apart is required to emplace or release the fuel pin (13b) such force being greater than the axial forces on the fuel pins (13b) which occur during operation of the reactor (12).

  7. Laser shockwave technique for characterization of nuclear fuel plate interfaces

    SciTech Connect

    Perton, M.; Levesque, D.; Monchalin, J.-P.; Lord, M.; Smith, J. A.; Rabin, B. H.

    2013-01-25

    The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

  8. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in...

  9. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in...

  10. SPENT NUCLEAR FUEL WASTE PACKAGE FILLER TESTING TECHNICAL GUIDELINES DOCUMENT

    SciTech Connect

    J.A. Cogar

    1996-08-28

    The purpose of this work is to provide supporting calculations for determination of the radiation source terms specific to subsurface shielding design and analysis. These calculations are not intended to provide the absolute values of the source terms, which are under the charter of the Waste Package Operations (WPO) Group. Rather, the calculations focus on evaluation of the various combinations of fuel enrichment, burnup and cooling time for a given decay heat output, consistent with the waste package (WP) thermal design basis. The objective is to determine the worst-case combination of the fuel characteristics (enrichment, burnup and cooling time) which would give the maximum radiation fields for subsurface shielding considerations. The calculations are limited to PWR fuel only, since the WP design is currently evolving with thinner walls and a reduced heat load as compared to the viability assessment (VA) reference design. The results for PWR fuel will provide a comparable indication of the trend for BWR fuel, as their characteristics are similar. The source term development for defense high-level waste and other spent nuclear fuel (SNF) is the responsibility of the WPO Group, and therefore, is not included this work. This work includes the following items responsive to the stated purpose and objective: (1) Determine the possible fuel parameters (initial enrichment, burnup and cooling time), that give the same decay heat value as specified for the waste package thermal design; (2) Obtain the neutron and gamma source terms for the various combinations of the fuel parameters for use in radiation field calculations; and (3) Calculate radiation fields on the surfaces of the waste package and its transporter to quantify the effects of the fuel parameters with the same decay heat value for use in identifying the worst-case combination of the fuel parameters.

  11. An approach for evaluating the integrity of fuel applied in Innovative Nuclear Energy Systems

    NASA Astrophysics Data System (ADS)

    Nakae, Nobuo; Ozawa, Takayuki; Ohta, Hirokazu; Ogata, Takanari; Sekimoto, Hiroshi

    2014-03-01

    One of the important issues in the study of Innovative Nuclear Energy Systems is evaluating the integrity of fuel applied in Innovative Nuclear Energy Systems. An approach for evaluating the integrity of the fuel is discussed here based on the procedure currently used in the integrity evaluation of fast reactor fuel. The fuel failure modes determining fuel life time were reviewed and fuel integrity was analyzed and compared with the failure criteria.

  12. Modern new nuclear fuel characteristics and radiation protection aspects.

    PubMed

    Terry, Ian R

    2005-01-01

    The glut of fissile material from reprocessing plants and from the conclusion of the cold war has provided the opportunity to design new fuel types to beneficially dispose of such stocks by generating useful power. Thus, in addition to the normal reactor core complement of enriched uranium fuel assemblies, two other types are available on the world market. These are the ERU (enriched recycled uranium) and the MOX (mixed oxide) fuel assemblies. Framatome ANP produces ERU fuel assemblies by taking feed material from reprocessing facilities and blending this with highly enriched uranium from other sources. MOX fuel assemblies contain plutonium isotopes, thus exploiting the higher neutron yield of the plutonium fission process. This paper describes and evaluates the gamma, spontaneous and alpha reaction neutron source terms of these non-irradiated fuel assembly types by defining their nuclear characteristics. The dose rates which arise from these terms are provided along with an overview of radiation protection aspects for consideration in transporting and delivering such fuel assemblies to power generating utilities.

  13. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    NASA Astrophysics Data System (ADS)

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-04-01

    The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. This approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.

  14. A European perspective on the US nuclear fuel cycle

    SciTech Connect

    Wilkinson, W.L.

    1989-01-01

    Many Europeans believe that the main problems which have impeded progress in solving the back end of the nuclear fuel cycle in the United States have been a series of ideological and political hang-ups and these, coupled with excessive bureaucracy, have made logical decision making on the back-end problems impossible. This situation has been caused by a succession of political nondecisions. Public confidence in nuclear generation was thereby undermined and, because of plentiful supplies of other energy sources, there was no urgent need to expand the nuclear program in the United States. Since uranium was cheap and fast reactors not commercially attractive, there was no economic incentive to reprocess fuel from existing reactors in the United States. The problem facing the United States is that of managing the large stocks of spent fuel which have arisen over many years. A logical way forward for the United States would appear to be as follows: build more storage for spent fuel; consider overseas reprocessing to provide plutonium; develop reprocessing technology; and develop direct disposal technology.

  15. New approaches to reprocessing of oxide nuclear fuel.

    PubMed

    Myasoedov, B F; Kulyako, Yu M

    Dissolution of UO2, U3O8, and solid solutions of actinides in UO2 in subacid aqueous solutions (pH 0.9-1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL(-1)). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.

  16. Advanced Nuclear Fuel Cycle Transitions: Optimization, Modeling Choices, and Disruptions

    NASA Astrophysics Data System (ADS)

    Carlsen, Robert W.

    Many nuclear fuel cycle simulators have evolved over time to help understan the nuclear industry/ecosystem at a macroscopic level. Cyclus is one of th first fuel cycle simulators to accommodate larger-scale analysis with it liberal open-source licensing and first-class Linux support. Cyclus also ha features that uniquely enable investigating the effects of modeling choices o fuel cycle simulators and scenarios. This work is divided into thre experiments focusing on optimization, effects of modeling choices, and fue cycle uncertainty. Effective optimization techniques are developed for automatically determinin desirable facility deployment schedules with Cyclus. A novel method fo mapping optimization variables to deployment schedules is developed. Thi allows relationships between reactor types and scenario constraints to b represented implicitly in the variable definitions enabling the usage o optimizers lacking constraint support. It also prevents wasting computationa resources evaluating infeasible deployment schedules. Deployed power capacit over time and deployment of non-reactor facilities are also included a optimization variables There are many fuel cycle simulators built with different combinations o modeling choices. Comparing results between them is often difficult. Cyclus flexibility allows comparing effects of many such modeling choices. Reacto refueling cycle synchronization and inter-facility competition among othe effects are compared in four cases each using combinations of fleet of individually modeled reactors with 1-month or 3-month time steps. There are noticeable differences in results for the different cases. The larges differences occur during periods of constrained reactor fuel availability This and similar work can help improve the quality of fuel cycle analysi generally There is significant uncertainty associated deploying new nuclear technologie such as time-frames for technology availability and the cost of buildin advanced reactors

  17. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    SciTech Connect

    J.K. Knudson

    2003-10-02

    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

  18. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    SciTech Connect

    A. Alsaed

    2005-07-28

    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

  19. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    SciTech Connect

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  20. Regulation of spent nuclear fuel shipment: A state perspective

    SciTech Connect

    Halstead, R.J.; Sinderbrand, C.; Woodbury, D.

    1987-01-01

    In 1985, the Wisconsin Department of Natural Resources (WDNR) sought to regulate rail shipments of spent nuclear fuel through the state, because federal regulations did not adequately protect the environmentally sensitive corridor along the route of the shipments. A state interagency working group identified five serious deficiencies in overall federal regulatory scheme: 1) failure to consider the safety or environmental risks associated with selected routes; 2) abscence of route-specific emergency response planning; 3) failure of the NRC to regulate the carrier of spent nuclear fuel or consider its safety record; 4) abscence of requirements for determination of need for, or the propriety of, specific shipments of spent nuclear fuel; and 5) the lack of any opportunity for meaningful public participation with respect to the decision to transport spent nuclear fuel. Pursuant to Wisconsin's hazardous substance statutes, the WDNR issues an order requiring the utility to file a spill prevention and mitigation plan or cease shipping through Wisconsin. A state trial court judge upheld the utility's challenge to Wisconsin's spill plan requirements, based on federal preemption of state authority. The state is now proposing federal legislation which would require: 1) NRC determination of need prior to approval of offsite shipment of spent fuel by the licensees; 2) NRC assessment of the potential environmental impacts of shipments along the proposed route, and comparative evaluation of alternative modes and routes; and 3) NRC approval of a route-specific emergency response and mitigation plan, including local training and periodic exercises. Additionally, the proposed legislation would authorize States and Indian Tribes to establish regulatory programs providing for permits, inspection, contingency plans for monitoring, containments, cleanup and decontamination, surveillance, enforcement and reasonable fees. 15 refs.

  1. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    NASA Technical Reports Server (NTRS)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  2. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    SciTech Connect

    Ewing, Rodney C.

    2003-09-14

    The successful disposal of spent nuclear fuel (SNF) is one of the most serious challenges to the successful completion of the nuclear fuel cycle and the future of nuclear power generation. In the United States, 21 percent of the electricity is generated by 107 commercial nuclear power plants (NPP), each of which generates 20 metric tons of spent nuclear fuel annually. In 1996, the total accumulation of spent nuclear fuel was 33,700 metric tons of heavy metal (MTHM) stored at 70 sites around the country. The end-of-life projection for current nuclear power plants (NPP) is approximately 86,000 MTHM. In the proposed nuclear waste repository at Yucca Mountain over 95% of the radioactivity originates from spent nuclear fuel. World-wide in 1998, approximately 130,000 MTHM of SNF have accumulated, most of it located at 236 NPP in 36 countries. Annual production of SNF is approximately 10,000 MTHM, containing about 100 tons of ''reactor grade'' plutonium. Any reasonable increase in the proportion of energy production by NPP, i.e., as a substitute for hydrocarbon-based sources of energy, will significantly increase spent nuclear fuel production. Spent nuclear fuel is essentially UO{sub 2} with approximately 4-5 atomic percent actinides and fission product elements. A number of these elements have long half-lives hence, the long-term behavior of the UO{sub 2} is an essential concern in the evaluation of the safety and risk of a repository for spent nuclear fuel. One of the unique and scientifically most difficult aspects of the successful disposal of spent nuclear fuel is the extrapolation of short-term laboratory data (hours to years) to the long time periods (10{sup 3} to 10{sup 5} years) as required by the performance objectives set in regulations, i.e. 10 CFR 60. The direct verification of these extrapolations or interpolations is not possible, but methods must be developed to demonstrate compliance with government regulations and to satisfy the public that there is a

  3. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    NASA Technical Reports Server (NTRS)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  4. A High Integrity Can Design for Degraded Nuclear Fuel

    SciTech Connect

    Holmes, P.A.

    1999-08-01

    A high integrity can (HIC), designed to meet the ASME Boiler and Pressure Vessel Code (Section III, Div. 3, static conditions) is proposed for the interim storage and repository disposal of Department of Energy (DOE) spent nuclear fuel. The HIC will be approximately 5 3/8 inches (134.38mm) in outside diameter with 1/4 inch (6.35mm) thick walls, and have a removable lid with a metallic seal that is capable of being welded shut. The opening of the can is approximately 4 3/8 inches (111.13mm). The HIC is primarily designed to contain items in the DOE SNF inventory that do not meet acceptance standards for direct disposal in a geologic repository. This includes fuel in the form of particulate dusts, sectioned pieces of fuel, core rubble, melted or degraded (non-intact) fuel elements, unclad uranium alloys, metallurgical specimens, and chemically reactive fuel components. The HIC is intended to act as a substitute cladding for the spent nuclear fuel, further isolate problematic materials, provide a long-term corrosion barrier, and add an extra internal pressure barrier to the waste package. The HIC will also delay potential fission product release and maintain geometry control for extended periods of time. For the entire disposal package to be licensed by the Nuclear Regulatory Commission, a HIC must effectively eliminate the disposal problems associated with problem SNF including the release of radioactive and/or reactive material and over pressurization of the HIC due to chemical reactions within the can. Two HICs were analyzed to envelop a range of can lengths between 42 and 101 inches. Using Abacus software, the HIC's were analyzed for end, side, and corner drops. Hastelloy C-22 was chosen based upon structural integrity, corrosion resistance, and neutron adsorption properties.

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

    SciTech Connect

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

    2010-11-01

    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.

  6. Nuclear reactor vessel fuel thermal insulating barrier

    DOEpatents

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    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.

  7. A combined gas cooled nuclear reactor and fuel cell cycle

    NASA Astrophysics Data System (ADS)

    Palmer, David J.

    Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

  8. Portable instrument for inspecting irradiated nuclear fuel assemblies

    DOEpatents

    Nicholson, Nicholas; Dowdy, Edward J.; Holt, David M.; Stump, Jr., Charles J.

    1985-01-01

    A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

  9. Uranyl peroxide enhanced nuclear fuel corrosion in seawater

    PubMed Central

    Armstrong, Christopher R.; Nyman, May; Shvareva, Tatiana; Sigmon, Ginger E.; Burns, Peter C.; Navrotsky, Alexandra

    2012-01-01

    The Fukushima-Daiichi nuclear accident brought together compromised irradiated fuel and large amounts of seawater in a high radiation field. Based on newly acquired thermochemical data for a series of uranyl peroxide compounds containing charge-balancing alkali cations, here we show that nanoscale cage clusters containing as many as 60 uranyl ions, bonded through peroxide and hydroxide bridges, are likely to form in solution or as precipitates under such conditions. These species will enhance the corrosion of the damaged fuel and, being thermodynamically stable and kinetically persistent in the absence of peroxide, they can potentially transport uranium over long distances. PMID:22308442

  10. Nuclear Magnetic Resonance Applications to Unconventional Fossil Fuel Resources

    NASA Astrophysics Data System (ADS)

    Kleinberg, R. L.; Leu, G.

    2008-12-01

    Technical and economic projections strongly suggest that fossil fuels will continue to play a dominant role in the global energy market through at least the mid twenty-first century. However, low-cost conventional oil and gas will be depleted in that time frame. Therefore new sources of energy will be needed. We discuss two relatively untapped unconventional fossil fuels: heavy oil and gas hydrate. In both cases, nuclear magnetic resonance plays a key role in appraising the resource and providing information needed for designing production processes.

  11. Fuel subassembly leak test chamber for a nuclear reactor

    DOEpatents

    Divona, Charles J.

    1978-04-04

    A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained.

  12. Nuclear reactor fuel element with vanadium getter on cladding

    DOEpatents

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

  13. Method for reprocessing and separating spent nuclear fuels

    DOEpatents

    Krikorian, Oscar H.; Grens, John Z.; Parrish, Sr., William H.

    1983-01-01

    Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

  14. Method for reprocessing and separating spent nuclear fuels. [Patent application

    DOEpatents

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

    1982-01-19

    Spent nuclear fuels, including actinide fuels, volatile and nonvolatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

  15. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  16. Mechanical Properties of Nuclear Fuel Surrogates using Picosecond Laser Ultrasonics

    SciTech Connect

    David Hurley; Marat Khafizov; Farhad Farzbod; Eric Burgett

    2013-05-01

    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

  17. Cell for a spent nuclear fuel rack

    SciTech Connect

    Flynn, W.M.

    1987-09-22

    This patent describes a cell for a spent fuel rack, comprising: a sheet metal element having an inside surface and an outside surface, and including a first flat wall portion, a second flat wall portion disposed perpendicularly to the first wall portion, a third flat wall portion disposed perpendicularly to the second wall portion and parallel to the first wall portion, a fourth flat wall portion disposed perpendicularly to the first and third wall portions and parallel to the second wall portion, an elongated bent region joining the first and second wall portions, an additional elongated bent region joining the second and third wall portions, a further elongated bent region joining the third and fourth wall portions, another elongated flat platform portion that is disposed parallel to the fourth wall portion but that is not coplanar with the fourth wall portion, and another elongated flat platform portion that is disposed parallel to the first wall portion but that is not coplanar with the first wall portion; means for joining the another platform portions; four sheets of neutron poison; and four sheet metal wrappers, each securing a respective neutron poison sheet to a respective wall portion.

  18. HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER

    SciTech Connect

    BROWN,LC; BESENBRUCH,GE; LENTSCH,RD; SCHULTZ,KR; FUNK,JF; PICKARD,PS; MARSHALL,AC; SHOWALTER,SK

    2003-06-01

    OAK B202 HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER. Combustion of fossil fuels, used to power transportation, generate electricity, heat homes and fuel industry provides 86% of the world's energy. Drawbacks to fossil fuel utilization include limited supply, pollution, and carbon dioxide emissions. Carbon dioxide emissions, thought to be responsible for global warming, are now the subject of international treaties. Together, these drawbacks argue for the replacement of fossil fuels with a less-polluting potentially renewable primary energy such as nuclear energy. Conventional nuclear plants readily generate electric power but fossil fuels are firmly entrenched in the transportation sector. Hydrogen is an environmentally attractive transportation fuel that has the potential to displace fossil fuels. Hydrogen will be particularly advantageous when coupled with fuel cells. Fuel cells have higher efficiency than conventional battery/internal combustion engine combinations and do not produce nitrogen oxides during low-temperature operation. Contemporary hydrogen production is primarily based on fossil fuels and most specifically on natural gas. When hydrogen is produced using energy derived from fossil fuels, there is little or no environmental advantage. There is currently no large scale, cost-effective, environmentally attractive hydrogen production process available for commercialization, nor has such a process been identified. The objective of this work is to find an economically feasible process for the production of hydrogen, by nuclear means, using an advanced high-temperature nuclear reactor as the primary energy source. Hydrogen production by thermochemical water-splitting (Appendix A), a chemical process that accomplishes the decomposition of water into hydrogen and oxygen using only heat or, in the case of a hybrid thermochemical process, by a combination of heat and electrolysis, could meet these goals. Hydrogen produced from fossil

  19. Changing Perspectives on Nonproliferation and Nuclear Fuel Cycles

    SciTech Connect

    Choi, J; Isaacs, T H

    2005-03-29

    The concepts of international control over technologies and materials in the proliferation sensitive parts of the nuclear fuel cycle, specifically those related to enrichment and reprocessing, have been the subject of many studies and initiatives over the years. For examples: the International Fissionable Material Storage proposal in President Eisenhower's Speech on Atoms for Peace, and in the Charter of the International Atomic Energy Agency (IAEA) when the organization was formed in 1957; the regional nuclear fuel cycle center centers proposed by INFCE in the 80's; and most recently and notably, proposals by Dr. ElBaradei, the Director General of IAEA to limit production and processing of nuclear weapons usable materials to facilities under multinational control; and by U.S. President George W. Bush, to limit enrichment and reprocessing to States that have already full scale, functioning plants. There are other recent proposals on this subject as well. In this paper, the similarities and differences, as well as the effectiveness and challenges in proliferation prevention of these proposals and concepts will be discussed. The intent is to articulate a ''new nuclear regime'' and to develop concrete steps to implement such regime for future nuclear energy and deployment.

  20. Effects of Water in Canisters Containing DOE Spent Nuclear Fuel

    SciTech Connect

    P. A. Lessing

    1998-10-01

    This document is a starting analysis in an attempt to quantify the potential of problems resulting from the presence of condensed water and water vapor in sealed storage canisters containing spent nuclear fuel (SNF). Of the 55 spent fuel types, it was determined that only three types required study to evaluate the effects of water, these types were represented by: (1) zirconium clad N-Reactor fuel, (2) aluminum clad Advanced Test Reactor (ATR) or Materials Test Reactor (MTR) fuel, (3) damaged and/or melted zirconium clad Three Mile Island (TMI) fuel. The spent fuel types were briefly described and potential problems were generalized into four catagories: (1) effects of water and the source(s) of that water, (2) effects of hydrogen generated from water, (3) pyrophoricity of uranium hydride, (4) swelling or cavitation. An extensive literature search was conducted, chemical reactions considered, and simple pressurization sensitivity calculations performed. Using the available information, conclusions were drawn and issues identified that need further consideration or analysis.

  1. Nuclear and materials aspects of the thorium fuel cycle

    NASA Astrophysics Data System (ADS)

    Rodriguez, P.; Sundaram, C. V.

    1981-09-01

    This paper is an attempt to assess and review the materials aspects of the thorium fuel cycle. It starts with an examination of the nuclear aspects of the thorium fuel cycle, meant as an introduction for materials scientists and engineers who may not normally be familiar with the concepts and terms involved. After defining and describing the thorium and uranium fuel cycles, the reasons for the resurgence of interest in the thorium fuel cycle and the technical and economic considerations that support its early adoption are examined. The reactor physics and fissile economics aspects of the thorium and uranium cycles are then compared. The specific reactor types suitable for the adoption of the thorium cycle are briefly examined and described. Subsequent sections of the paper are devoted to a detailed discussion of the materials aspects of the thorium fuel cycle. Available information on fabrication, refabrication and irradiation performance of thorium-based fuels for light water reactors, heavy water reactors, high temperature gas-cooled reactors, molten salt breeder reactors and fast breeder reactors is critically reviewed and analysed. Materials problems related to cladding and structural materials are also discussed whenever these are unique to the thorium cycle.

  2. Molten tin reprocessing of spent nuclear fuel elements

    DOEpatents

    Heckman, Richard A.

    1983-01-01

    A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

  3. Financing Strategies For A Nuclear Fuel Cycle Facility

    SciTech Connect

    David Shropshire; Sharon Chandler

    2006-07-01

    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

  4. Method for cleaning solution used in nuclear fuel reprocessing

    DOEpatents

    Tallent, Othar K.; Crouse, David J.; Mailen, James C.

    1982-01-01

    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.

  5. Method for cleaning solution used in nuclear fuel reprocessing

    DOEpatents

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

    1980-12-17

    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.

  6. Letter Report: Looking Ahead at Nuclear Fuel Resources

    SciTech Connect

    J. Stephen Herring

    2013-09-01

    The future of nuclear energy and its ability to fulfill part of the world’s energy needs for centuries to come depend on a reliable input of nuclear fuel, either thorium or uranium. Obviously, the present nuclear fuel cycle is completely dependent on uranium. Future thorium cycles will also depend on 235U or fissile isotopes separated from used fuel to breed 232Th into fissile 233U. This letter report discusses several emerging areas of scientific understanding and technology development that will clarify and enable assured supplies of uranium and thorium well into the future. At the most fundamental level, the nuclear energy community needs to appreciate the origins of uranium and thorium and the processes of planetary accretion by which those materials have coalesced to form the earth and other planets. Secondly, the studies of geophysics and geochemistry are increasing understanding of the processes by which uranium and thorium are concentrated in various locations in the earth’s crust. Thirdly, the study of neutrinos and particularly geoneutrinos (neutrinos emitted by radioactive materials within the earth) has given an indication of the overall global inventories of uranium and thorium, though little indication for those materials’ locations. Crustal temperature measurements have also given hints of the vertical distribution of radioactive heat sources, primarily 238U and 232Th, within the continental crust. Finally, the evolving technologies for laser isotope separation are indicating methods for reducing the energy input to uranium enrichment but also for tailoring the isotopic vectors of fuels, burnable poisons and structural materials, thereby adding another tool for dealing with long-term waste management.

  7. Evaluation of a potential nuclear fuel repository criticality: Lessons learned

    SciTech Connect

    Wilson, J.R.; Evans, D.

    1995-10-01

    This paper presents lessons learned from a Probabilistic Risk Assessment (PRA) of the potential for a criticality in a repository containing spent nuclear fuel with high enriched uranium. The insights gained consisted of remarkably detailed conclusions about design issues, failure mechanisms, frequencies and source terms for events up to 10,000 years in the future. Also discussed are the approaches taken by the analysts in presenting this very technical report to a nontechnical and possibly antagonistic audience.

  8. Naval Spent Nuclear Fuel disposal Container System Description Document

    SciTech Connect

    N. E. Pettit

    2001-07-13

    The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the natural barrier will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

  9. Spent nuclear fuel canister storage building conceptual design report

    SciTech Connect

    Swenson, C.E.

    1996-01-01

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

  10. Asymptotic Expansion Homogenization for Multiscale Nuclear Fuel Analysis

    SciTech Connect

    Hales, J. D.; Tonks, M. R.; Chockalingam, K.; Perez, D. M.; Novascone, S. R.; Spencer, B. W.; Williamson, R. L.

    2015-03-01

    Engineering scale nuclear fuel performance simulations can benefit by utilizing high-fidelity models running at a lower length scale. Lower length-scale models provide a detailed view of the material behavior that is used to determine the average material response at the macroscale. These lower length-scale calculations may provide insight into material behavior where experimental data is sparse or nonexistent. This multiscale approach is especially useful in the nuclear field, since irradiation experiments are difficult and expensive to conduct. The lower length-scale models complement the experiments by influencing the types of experiments required and by reducing the total number of experiments needed. This multiscale modeling approach is a central motivation in the development of the BISON-MARMOT fuel performance codes at Idaho National Laboratory. These codes seek to provide more accurate and predictive solutions for nuclear fuel behavior. One critical aspect of multiscale modeling is the ability to extract the relevant information from the lower length-scale sim- ulations. One approach, the asymptotic expansion homogenization (AEH) technique, has proven to be an effective method for determining homogenized material parameters. The AEH technique prescribes a system of equations to solve at the microscale that are used to compute homogenized material constants for use at the engineering scale. In this work, we employ AEH to explore the effect of evolving microstructural thermal conductivity and elastic constants on nuclear fuel performance. We show that the AEH approach fits cleanly into the BISON and MARMOT codes and provides a natural, multidimensional homogenization capability.

  11. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1986-01-01

    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.

  12. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.

    1983-08-29

    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 engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  13. Vision for the development of an international nuclear fuel recycling program

    SciTech Connect

    Kok, Kenneth D.

    2007-07-01

    The purpose of the development of an international nuclear fuel recycle program is to: - Demonstrate advanced recycling by working to prove the technologies needed to close the fuel cycle, minimize waste, and obtain more energy benefit for each unit of fuel. - Build a global vision by enlisting partners to limit the spread of sensitive nuclear technologies in a way that enables nuclear power to meet global challenges. The program will begin with the establishment of a smaller scale secure fuel cycle facility that would serve as a model for international nuclear fuel reprocessing centers that would eventually be built in several countries world wide. The operating process plants will provide the secure and safe guarded environment for the recycle of spent fuel from nuclear power stations around the world. The demonstration site will provide for developing and testing processes that would lead to the more complete use of the energy available in nuclear fuels and the minimization of long lived nuclear waste. (author)

  14. NUCLEAR-FUELED CIRCULATORY SUPPORT SYSTEMS IV: RADIOLOGIC PERSPECTIVES

    PubMed Central

    Huffman, F. N.; Norman, J. C.

    1974-01-01

    If an implantable artificial heart can be developed, it should prove beneficial to a significant group of patients. A variety of energy sources, such as biologic, electromagnetic, and nuclear, are under evaluation. Currently, biologic fuel cell technology is not sufficiently advanced to permit its extrapolation to the power levels required for implantable circulatory support systems. Electromagnetic systems have the disadvantage of heavy batteries of considerable bulk requiring frequent recharging. Radioisotope-fueled thermal engine systems have the potential of providing degrees of freedom not possible with rechargeable units. However, radiosotope circulatory support systems subject their recipients to prolonged intracorporeal radiation, add to environmental background radiation, and constitute an exceedingly small, but finite, hazard due to possible violation of fuel containment. PMID:15215965

  15. Thermal barrier and support for nuclear reactor fuel core

    DOEpatents

    Betts, Jr., William S.; Pickering, J. Larry; Black, William E.

    1987-01-01

    A thermal barrier/core support for the fuel core of a nuclear reactor having a metallic cylinder secured to the reactor vessel liner and surrounded by fibrous insulation material. A top cap is secured to the upper end of the metallic cylinder that locates and orients a cover block and post seat. Under normal operating conditions, the metallic cylinder supports the entire load exerted by its associated fuel core post. Disposed within the metallic cylinder is a column of ceramic material, the height of which is less than that of the metallic cylinder, and thus is not normally load bearing. In the event of a temperature excursion beyond the design limits of the metallic cylinder and resulting in deformation of the cylinder, the ceramic column will abut the top cap to support the fuel core post.

  16. Nuclear fuel spacer grid with improved outer straps

    SciTech Connect

    De Mario, E.E.; Knott, R.P.

    1986-04-29

    A grid is described for the spacing of fuel rods in a nuclear reactor fuel assembly, comprising: a plurality of interleaved grid straps arranged in an egg-crate configuration defining cells therein for the separate enclosure of each of the fuel rods; and four outer straps attached together in a generally square-shaped array surrounding the heightwise edges of the grid straps. The outer straps each has a central portion and a top and a bottom resilient lengthwise border portion, with the heightwise edges of the grid straps attached to the central portions of their associated outer straps and with the border portions extending vertically beyond and projecting horizontally outwardly beyond their associated central portions of their common outer straps.

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

    SciTech Connect

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

    2014-01-09

    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

  18. A physical and economic model of the nuclear fuel cycle

    NASA Astrophysics Data System (ADS)

    Schneider, Erich Alfred

    A model of the nuclear fuel cycle that is suitable for use in strategic planning and economic forecasting is presented. The model, to be made available as a stand-alone software package, requires only a small set of fuel cycle and reactor specific input parameters. Critical design criteria include ease of use by nonspecialists, suppression of errors to within a range dictated by unit cost uncertainties, and limitation of runtime to under one minute on a typical desktop computer. Collision probability approximations to the neutron transport equation that lead to a computationally efficient decoupling of the spatial and energy variables are presented and implemented. The energy dependent flux, governed by coupled integral equations, is treated by multigroup or continuous thermalization methods. The model's output includes a comprehensive nuclear materials flowchart that begins with ore requirements, calculates the buildup of 24 actinides as well as fission products, and concludes with spent fuel or reprocessed material composition. The costs, direct and hidden, of the fuel cycle under study are also computed. In addition to direct disposal and plutonium recycling strategies in current use, the model addresses hypothetical cycles. These include cycles chosen for minor actinide burning and for their low weapons-usable content.

  19. Classification of spent reactor fuel for nuclear forensics.

    PubMed

    Jones, Andrew E; Turner, Phillip; Zimmerman, Colin; Goulermas, John Y

    2014-06-03

    In this paper we demonstrate the use of pattern recognition and machine learning techniques to determine the reactor type from which spent reactor fuel has originated. This has been done using the isotopic and elemental measurements of the sample and proves to be very useful in the field of nuclear forensics. Nuclear materials contain many variables (impurities and isotopes) that are very difficult to consider individually. A method that considers all material parameters simultaneously is advantageous. Currently the field of nuclear forensics focuses on the analysis of key material properties to determine details about the materials processing history, for example, utilizing known half-lives of isotopes can determine when the material was last processed (Stanley, F. E. J. Anal. At. Spectrom. 2012, 27, 1821; Varga, Z.; Wallenius, M.; Mayer, K.; Keegan, E.; Millet, S. Anal. Chem. 2009, 81, 8327-8334). However, it has been demonstrated that multivariate statistical analysis of isotopic concentrations can complement these method and are able to make use of a greater level of information through dimensionality reduction techniques (Robel, M.; Kristo, M. J. J. Environ. Radioact. 2008, 99, 1789-1797; Robel, M.; Kristo, M. J.; Heller, M. A. Nuclear Forensic Inferences Using Iterative Multidimensional Statistics. In Proceedings of the Institute of Nuclear Materials Management 50th Annual Meeting, Tucson, AZ, July 2009; 12 pages; Nicolaou, G. J. Environ. Radioact. 2006, 86, 313-318; Pajo, L.; Mayer, K.; Koch, L. Fresenius' J. Anal. Chem. 2001, 371, 348-352). There has been some success in using such multidimensional statistical methods to determine details about the history of spent reactor fuel (Robel, M.; Kristo, M. J. J. Environ. Radioact. 2008, 99, 1789-1797). Here, we aim to expand on these findings by pursuing more robust dimensionality reduction techniques based on manifold embedding which are able to better capture the intrinsic data set information. Furthermore, we

  20. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  1. National briefing summaries: Nuclear fuel cycle and waste management

    SciTech Connect

    Schneider, K.J.; Bradley, D.J.; Fletcher, J.F.; Konzek, G.J.; Lakey, L.T.; Mitchell, S.J.; Molton, P.M.; Nightingale, R.E.

    1991-04-01

    Since 1976, the International Program Support Office (IPSO) at the Pacific Northwest Laboratory (PNL) has collected and compiled publicly available information concerning foreign and international radioactive waste management programs. This National Briefing Summaries is a printout of an electronic database that has been compiled and is maintained by the IPSO staff. The database contains current information concerning the radioactive waste management programs (with supporting information on nuclear power and the nuclear fuel cycle) of most of the nations (except eastern European countries) that now have or are contemplating nuclear power, and of the multinational agencies that are active in radioactive waste management. Information in this document is included for three additional countries (China, Mexico, and USSR) compared to the prior issue. The database and this document were developed in response to needs of the US Department of Energy.

  2. Managing the Nuclear Fuel Cycle, The Big Picture

    SciTech Connect

    Brett W Carlsen

    2010-07-01

    The nuclear industry, at least in the United States, has failed to deliver on its promise of cheap, abundant energy. After pioneering the science and application and becoming a primary exporter of nuclear technologies, domestic use of nuclear power fell out-of-favor with the public and has been relatively stagnant for several decades. Recently, renewed interest has generated optimism and talk of a nuclear renaissance characterized by a new generation of safe, clean nuclear plants in this country. But, as illustrated by recent policy shifts regarding closure of the fuel cycle and geologic disposal of high-level radioactive wastes, significant hurdles have yet to be overcome. Using the principles of system dynamics, this paper will take a holistic look at the nuclear industry and the interactions between the key players to explore both the intended and unintended consequences of efforts to address the issues that have impeded the growth of the industry and also to illustrate aspects which must be effectively addressed if the renaissance of our industry is to be achieved and sustained.

  3. Methodology for estimating reprocessing costs for nuclear fuels

    SciTech Connect

    Carter, W. L.; Rainey, R. H.

    1980-02-01

    A technological and economic evaluation of reprocessing requirements for alternate fuel cycles requires a common assessment method and a common basis to which various cycles can be related. A methodology is described for the assessment of alternate fuel cycles utilizing a side-by-side comparison of functional flow diagrams of major areas of the reprocessing plant with corresponding diagrams of the well-developed Purex process as installed in the Barnwell Nuclear Fuel Plant (BNFP). The BNFP treats 1500 metric tons of uranium per year (MTU/yr). Complexity and capacity factors are determined for adjusting the estimated facility and equipment costs of BNFP to determine the corresponding costs for the alternate fuel cycle. Costs of capacities other than the reference 1500 MT of heavy metal per year are estimated by the use of scaling factors. Unit costs of reprocessed fuel are calculated using a discounted cash flow analysis for three economic bases to show the effect of low-risk, typical, and high-risk financing methods.

  4. VISION -- A Dynamic Model of the Nuclear Fuel Cycle

    SciTech Connect

    J. J. Jacobson; A. M. Yacout; S. J. Piet; D. E. Shropshire; G. E. Matthern

    2006-02-01

    The 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 deploy¬ment 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. VISION is intended to serve as a broad systems analysis and study tool applicable to work conducted as part of the AFCI (including costs estimates) and Generation IV reactor development studies.

  5. DOE-owned spent nuclear fuel program plan

    SciTech Connect

    1995-11-01

    The Department of Energy (DOE) has produced spent nuclear fuel (SNF) for many years as part of its various missions and programs. The historical process for managing this SNF was to reprocess it whereby valuable material such as uranium or plutonium was chemically separated from the wastes. These fuels were not intended for long-term storage. As the need for uranium and plutonium decreased, it became necessary to store the SNF for extended lengths of time. This necessity resulted from a 1992 DOE decision to discontinue reprocessing SNF to recover strategic materials (although limited processing of SNF to meet repository acceptance criteria remains under consideration, no plutonium or uranium extraction for other uses is planned). Both the facilities used for storage, and the fuel itself, began experiencing aging from this extended storage. New efforts are now necessary to assure suitable fuel and facility management until long-term decisions for spent fuel disposition are made and implemented. The Program Plan consists of 14 sections as follows: Sections 2--6 describe objectives, management, the work plan, the work breakdown structure, and the responsibility assignment matrix. Sections 7--9 describe the program summary schedules, site logic diagram, SNF Program resource and support requirements. Sections 10--14 present various supplemental management requirements and quality assurance guidelines.

  6. Optimization of a Dry, Mixed Nuclear Fuel Storage Array for Nuclear Criticality Safety

    NASA Astrophysics Data System (ADS)

    Baranko, Benjamin T.

    A dry storage array of used nuclear fuel at the Idaho National Laboratory contains a mixture of more than twenty different research and test reactor fuel types in up to 636 fuel storage canisters. New analysis demonstrates that the current arrangement of the different fuel-type canisters does not minimize the system neutron multiplication factor (keff), and that the entire facility storage capacity cannot be utilized without exceeding the subcritical limit (ksafe) for ensuring nuclear criticality safety. This work determines a more optimal arrangement of the stored fuels with a goal to minimize the system keff, but with a minimum of potential fuel canister relocation movements. The solution to this multiple-objective optimization problem will allow for both an improvement in the facility utilization while also offering an enhancement in the safety margin. The solution method applies stochastic approximation and a Tabu search metaheuristic to an empirical model developed from supporting MCNP calculations. The results establish an optimal relocation of between four to sixty canisters, which will allow the current thirty-one empty canisters to be used for storage while reducing the array keff by up to 0.018 +/- 0.003 relative to the current arrangement.

  7. Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

    NASA Astrophysics Data System (ADS)

    Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

    2016-08-01

    The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

  8. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    DOE PAGES

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-01-17

    The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

  9. MicroRaman Measurements for Nuclear Fuel Reprocessing Applications

    SciTech Connect

    Casella, Amanda; Lines, Amanda; Nelson, Gilbert; Bello, Job; Bryan, Samuel

    2016-01-01

    Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper will focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. This paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO3, and UO2(NO3)2.

  10. MicroRaman measurements for nuclear fuel reprocessing applications

    DOE PAGES

    Casella, Amanda; Lines, Amanda; Nelson, Gilbert; ...

    2016-12-01

    Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper willmore » focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. Finally, this paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO3, and UO2(NO3)2.« less

  11. MicroRaman measurements for nuclear fuel reprocessing applications

    SciTech Connect

    Casella, Amanda; Lines, Amanda; Nelson, Gilbert; Bello, Job; Bryan, Samuel

    2016-12-01

    Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper will focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. Finally, this paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO3, and UO2(NO3)2.

  12. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    NASA Astrophysics Data System (ADS)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  13. Future nuclear fuel cycles: Prospect and challenges for actinide recycling

    NASA Astrophysics Data System (ADS)

    Warin, Dominique

    2010-03-01

    The global energy context pleads in favour of a sustainable development of nuclear energy since the demand for energy will likely increase, whereas resources will tend to get scarcer and the prospect of global warming will drive down the consumption of fossil fuel. In this context, nuclear power has the worldwide potential to curtail the dependence on fossil fuels and thereby to reduce the amount of greenhouse gas emissions while promoting energy independence. How we deal with nuclear radioactive waste is crucial in this context. In France, the public's concern regarding the long-term waste management made the French Governments to prepare and pass the 1991 and 2006 Acts, requesting in particular the study of applicable solutions for still minimizing the quantity and the hazardousness of final waste. This necessitates High Active Long Life element (such as the Minor Actinides MA) recycling, since the results of fuel cycle R&D could significantly change the challenges for the storage of nuclear waste. HALL recycling can reduce the heat load and the half-life of most of the waste to be buried to a couple of hundred years, overcoming the concerns of the public related to the long-life of the waste and thus aiding the "burying approach" in securing a "broadly agreed political consensus" of waste disposal in a geological repository. This paper presents an overview of the recent R and D results obtained at the CEA Atalante facility on innovative actinide partitioning hydrometallurgical processes. For americium and curium partitioning, these results concern improvements and possible simplifications of the Diamex-Sanex process, whose technical feasibility was already demonstrated in 2005. Results on the first tests of the Ganex process (grouped actinide separation for homogeneous recycling) are also discussed. In the coming years, next steps will involve both better in-depth understanding of the basis of these actinide partitioning processes and, for the new promising

  14. Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

    SciTech Connect

    Unal, Cetin; Pasamehmetoglu, Kemal; Carmack, Jon

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

  15. Ablation study of tungsten-based nuclear thermal rocket fuel

    NASA Astrophysics Data System (ADS)

    Smith, Tabitha Elizabeth Rose

    The research described in this thesis has been performed in order to support the materials research and development efforts of NASA Marshall Space Flight Center (MSFC), of Tungsten-based Nuclear Thermal Rocket (NTR) fuel. The NTR was developed to a point of flight readiness nearly six decades ago and has been undergoing gradual modification and upgrading since then. Due to the simplicity in design of the NTR, and also in the modernization of the materials fabrication processes of nuclear fuel since the 1960's, the fuel of the NTR has been upgraded continuously. Tungsten-based fuel is of great interest to the NTR community, seeking to determine its advantages over the Carbide-based fuel of the previous NTR programs. The materials development and fabrication process contains failure testing, which is currently being conducted at MSFC in the form of heating the material externally and internally to replicate operation within the nuclear reactor of the NTR, such as with hot gas and RF coils. In order to expand on these efforts, experiments and computational studies of Tungsten and a Tungsten Zirconium Oxide sample provided by NASA have been conducted for this dissertation within a plasma arc-jet, meant to induce ablation on the material. Mathematical analysis was also conducted, for purposes of verifying experiments and making predictions. The computational method utilizes Anisimov's kinetic method of plasma ablation, including a thermal conduction parameter from the Chapman Enskog expansion of the Maxwell Boltzmann equations, and has been modified to include a tangential velocity component. Experimental data matches that of the computational data, in which plasma ablation at an angle shows nearly half the ablation of plasma ablation at no angle. Fuel failure analysis of two NASA samples post-testing was conducted, and suggestions have been made for future materials fabrication processes. These studies, including the computational kinetic model at an angle and the

  16. EARTHQUAKE CAUSED RELEASES FROM A NUCLEAR FUEL CYCLE FACILITY

    SciTech Connect

    Charles W. Solbrig; Chad Pope; Jason Andrus

    2014-08-01

    The fuel cycle facility (FCF) at the Idaho National Laboratory is a nuclear facility which must be licensed in order to operate. A safety analysis is required for a license. This paper describes the analysis of the Design Basis Accident for this facility. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. The hot cell is used to process spent metallic nuclear fuel. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities.

  17. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    SciTech Connect

    Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy; Scaglione, John M.

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  18. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    SciTech Connect

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

    2013-10-01

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  19. CESAR: A Code for Nuclear Fuel and Waste Characterisation

    SciTech Connect

    Vidal, J.M.; Grouiller, J.P.; Launay, A.; Berthion, Y.; Marc, A.; Toubon, H.

    2006-07-01

    CESAR (Simplified Evolution Code Applied to Reprocessing) is a depletion code developed through a joint program between CEA and COGEMA. In the late 1980's, the first use of this code dealt with nuclear measurement at the Laboratories of the La Hague reprocessing plant. The use of CESAR was then extended to characterizations of all entrance materials and for characterisation, via tracer, of all produced waste. The code can distinguish more than 100 heavy nuclides, 200 fission products and 100 activation products, and it can characterise both the fuel and the structural material of the fuel. CESAR can also make depletion calculations from 3 months to 1 million years of cooling time. Between 2003-2005, the 5. version of the code was developed. The modifications were related to the harmonisation of the code's nuclear data with the JEF2.2 nuclear data file. This paper describes the code and explains the extensive use of this code at the La Hague reprocessing plant and also for prospective studies. The second part focuses on the modifications of the latest version, and describes the application field and the qualification of the code. Many companies and the IAEA use CESAR today. CESAR offers a Graphical User Interface, which is very user-friendly. (authors)

  20. Hanford K Basins spent nuclear fuels project update

    SciTech Connect

    Hudson, F.G.

    1997-10-17

    Twenty one hundred metric tons of spent nuclear fuel are stored in two concrete pools on the Hanford Site, known as the K Basins, near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current wet pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in the K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported into the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building (CSB) in the 200 Area for staging prior to hot conditioning. The conditioning step to remove chemically bound water is performed by holding the MCO at 300 C under vacuum. This step is necessary to prevent excessive pressure buildup during interim storage that could be caused by corrosion. After conditioning, MCOs will remain in the CSB for interim storage until a national repository is completed.

  1. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    SciTech Connect

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-06

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  2. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-01

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  3. Uranium in the Nuclear Fuel Cycle: Creation of Plutonium (Invited)

    NASA Astrophysics Data System (ADS)

    Ewing, R. C.

    2009-12-01

    One of the important properties of uranium is that it can be used to “breed” higher actinides, particularly plutonium. During the past sixty years, more than 1,800 metric tonnes of Pu, and substantial quantities of the “minor” actinides, such as Np, Am and Cm, have been generated in nuclear reactors - a permanent record of nuclear power. Some of these transuranium elements can be a source of energy in fission reactions (e.g., 239Pu), a source of fissile material for nuclear weapons (e.g., 239Pu and 237Np), and of environmental concern because of their long-half lives and radiotoxicity (e.g., 239Pu and 237Np). In fact, the new strategies of the Advance Fuel Cycle Initiative (AFCI) are, in part, motivated by an effort to mitigate some of the challenges of the disposal of these long-lived actinides. There are two basic strategies for the disposition of these heavy elements: 1.) to “burn” or transmute the actinides using nuclear reactors or accelerators; 2.) to “sequester” the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. There has been substantial interest in the use of actinide-bearing minerals, such as zircon or isometric pyrochlore, A2B2O7 (A= rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage1. The radiation stability of these compositions is closely related to the structural distortions that can be accommodated for specific pyrochlore compositions and the electronic structure of the B-site cation. Recent developments in the understanding of the properties of heavy element solids have opened up new possibilities for the design of advanced nuclear fuels and waste forms.

  4. In-Pile Thermal Conductivity Measurement Method for Nuclear Fuels

    SciTech Connect

    Joy L. Rempe; Brandon Fox; Heng Ban; Joshua E. Daw; Darrell L. Knudson; Keith G. Condie

    2009-08-01

    Thermophysical properties of advanced nuclear fuels and materials during irradiation must be known prior to their use in existing, advanced, or next generation reactors. Thermal conductivity is one of the most important properties for predicting fuel and material performance. A joint Utah State University (USU) / Idaho National Laboratory (INL) project, which is being conducted with assistance from the Institute for Energy Technology at the Norway Halden Reactor Project, is investigating in-pile fuel thermal conductivity measurement methods. This paper focuses on one of these methods – a multiple thermocouple method. This two-thermocouple method uses a surrogate fuel rod with Joule heating to simulate volumetric heat generation to gain insights about in-pile detection of thermal conductivity. Preliminary results indicated that this method can measure thermal conductivity over a specific temperature range. This paper reports the thermal conductivity values obtained by this technique and compares these values with thermal property data obtained from standard thermal property measurement techniques available at INL’s High Test Temperature Laboratory. Experimental results and material properties data are also compared to finite element analysis results.

  5. Interatomic potentials for mixed oxide and advanced nuclear fuels

    SciTech Connect

    Tiwary, Pratyush; Walle, Axel van de; Jeon, Byoungseon; Groenbech-Jensen, Niels

    2011-03-01

    We extend our recently developed interatomic potentials for UO{sub 2} to the fuel system (U,Pu,Np)O{sub 2}. We do so by fitting against an extensive database of ab initio results as well as to experimental measurements. The applicability of these interactions to a variety of mixed environments beyond the fitting domain is also assessed. The employed formalism makes these potentials applicable across all interatomic distances without the need for any ambiguous splining to the well-established short-range Ziegler-Biersack-Littmark universal pair potential. We therefore expect these to be reliable potentials for carrying out damage simulations (and molecular dynamics simulations in general) in nuclear fuels of varying compositions for all relevant atomic collision energies.

  6. Laser cutting apparatus for nuclear core fuel subassembly

    DOEpatents

    Walch, Allan P.; Caruolo, Antonio B.

    1982-02-23

    The object of the invention is to provide a system and apparatus which employs laser cutting to disassemble a nuclear core fuel subassembly. The apparatus includes a gantry frame (C) which straddles the core fuel subassembly (14), an x-carriage (22) travelling longitudinally above the frame which carries a focus head assembly (D) having a vertically moving carriage (46) and a laterally moving carriage (52), a system of laser beam transferring and focusing mirrors carried by the x-carriage and focusing head assembly, and a shroud follower (F) and longitudinal follower (G) for following the shape of shroud (14) to maintain a beam focal point (44) fixed upon the shroud surface for accurate cutting.

  7. Separation of the rare-earth fission product poisons from spent nuclear fuel

    SciTech Connect

    Christian, Jerry D.; Sterbentz, James W.

    2016-08-30

    A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

  8. Spent Nuclear Fuel Project (SNFP) gas generation from N-Fuel in multi-canister overpacks

    SciTech Connect

    Cooper, T.D.

    1996-08-01

    During the conversion from wet pool storage for spent nuclear fuel at Hanford, gases will be generated from both radiolysis and chemical reactions. The gas generation phenomenon needs to be understood as it applies to safety and design issues,specifically over pressurization of sealed storage containers,and detonation/deflagration of flammable gases. This study provides an initial basis to predict the implications of gas generation on the proposed functional processes for spent nuclear fuel conversion from wet to dry storage. These projections are based upon examination of the history of fuel manufacture at Hanford, irradiation in the reactors, corrosion during wet pool storage, available fuel characterization data and available information from literature. Gas generation via radiolysis and metal corrosion are addressed. The study examines gas generation, the boundary conditions for low medium and high levels of sludge in SNF storage/processing containers. The functional areas examined include: flooded and drained Multi-Canister Overpacks, cold vacuum drying, shipping and staging and long term storage.

  9. Development of nuclear fuel microsphere handling techniques and equipment

    SciTech Connect

    Mack, J.E.; Suchomel, R.R.; Angelini, P.

    1980-01-01

    Considerable progress has been made in the development of microsphere handling techniques and equipment for nuclear applications. Work at Oak Ridge National Laboratory with microspherical fuel forms dates back to the early sixties with the development of the sol-gel process. Since that time a number of equipment items and systems specifically related to microsphere handling and characterization have been identified and developed for eventual application in a remote recycle facility. These include positive and negative pressure transfer systems, samplers, weighers, a blender-dispenser, and automated devices for particle size distribution and crushing strength analysis. The current status of these and other components and systems is discussed.

  10. Closure Mechanism and Method for Spent Nuclear Fuel Canisters

    SciTech Connect

    Doman, Marvin J.

    2004-11-23

    A canister is provided for storing, transporting, and/or disposing of spent nuclear fuel. The canister includes a canister shell, a top shield plug disposed within the canister, and a leak-tight closure arrangement. The closure arrangement includes a shear ring which forms a containment boundary of the canister, and which is welded to the canister shell and top shield plug. An outer seal plate, forming an outer seal, is disposed above the shear ring and is welded to the shield plug and the canister.

  11. An americium-fueled gas core nuclear rocket

    SciTech Connect

    Kammash, T.; Galbraith, D.L.; Jan, T. )

    1993-01-10

    A gas core fission reactor that utilizes americium in place of uranium is examined for potential utilization as a nuclear rocket for space propulsion. The isomer [sup 242m]Am with a half life of 141 years is obtained from an (n, [gamma]) capture reaction with [sup 241]Am, and has the highest known thermal fission cross section. We consider a 7500 MW reactor, whose propulsion characteristics with [sup 235]U have already been established, and re-examine it using americium. We find that the same performance can be achieved at a comparable fuel density, and a radial size reduction (of both core and moderator/reflector) of about 70%.

  12. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    SciTech Connect

    Rodney C. Ewing

    2004-10-07

    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.

  13. Process for recovery of palladium from nuclear fuel reprocessing wastes

    DOEpatents

    Campbell, David O.; Buxton, Samuel R.

    1981-01-01

    Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M, (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound, (c) heating the solution at reflux temperature until precipitation is complete, and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

  14. Process for recovery of palladium from nuclear fuel reprocessing wastes

    DOEpatents

    Campbell, D.O.; Buxton, S.R.

    1980-06-16

    Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M; (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound; (c) heating the solution at reflux temperature until precipitation is complete; and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

  15. Automatic inspection system for nuclear fuel pellets or rods

    DOEpatents

    Miller, Jr., William H.; Sease, John D.; Hamel, William R.; Bradley, Ronnie A.

    1978-01-01

    An automatic inspection system is provided for determining surface defects on cylindrical objects such as nuclear fuel pellets or rods. The active element of the system is a compound ring having a plurality of pneumatic jet units directed into a central bore. These jet units are connected to provide multiple circuits, each circuit being provided with a pressure sensor. The outputs of the sensors are fed to a comparator circuit whereby a signal is generated when the difference of pressure between pneumatic circuits, caused by a defect, exceeds a pre-set amount. This signal may be used to divert the piece being inspected into a "reject" storage bin or the like.

  16. Method for cleaning solution used in nuclear fuel reprocessing

    DOEpatents

    Tallent, Othar K.; Dodson, Karen E.; Mailen, James C.

    1983-01-01

    A nuclear fuel processing solution containing (1) hydrocarbon diluent, (2) tri-n-butyl phosphate or tri-2-ethylhexyl phosphate, and (3) monobutyl phosphate, dibutyl phosphate, mono-2-ethylhexyl phosphate, di-2-ethylhexyl phosphate, or a complex formed by plutonium, uranium, or a fission product thereof with monobutyl phosphate, dibutyl phosphate, mono-2-ethylhexyl phosphate, or di-2-ethylhexyl phosphate is contacted with silica gel having alkali ions absorbed thereon to remove any one of the degradation products named in section (3) above from said solution.

  17. Nuclear fuel element with axially aligned fuel pellets and fuel microspheres therein

    DOEpatents

    Sease, J.D.; Harrington, F.E.

    1973-12-11

    Elongated single- and multi-region fuel elements are prepared by replacing within a cladding container a coarse fraction of fuel material which includes plutonium and uranium in the appropriate regions of the fuel element and then infiltrating with vibration a fine-sized fraction of uranium-containing microspheres throughout all interstices in the coarse material in a single loading. The fine, rigid material defines a thin annular layer between the coarse fraction and the cladding to reduce adverse mechanical and chemical interactions. (Official Gazette)

  18. Fuel leak detection apparatus for gas cooled nuclear reactors

    DOEpatents

    Burnette, Richard D.

    1977-01-01

    Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

  19. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2007-11-01

    pounds of U3O8 were purchased for U.S. nuclear power reactors in 2006, of which 16% was U.S. origin.28 The balance was made up in part by imports and...enriching nearly 800,000 lbs. of uranium “yellowcake” ( U3O8 ). Approximately 102 Figure 1. The Conceptual Nuclear Fuel Cycle CRS-12 29 U.S. DOE, Report... U3O8 . 29 The International Atomic Energy Agency (IAEA) projects that the demand for uranium will begin to exceed supply after 2010, and by as much as

  20. MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY

    SciTech Connect

    Hill, Thomas J

    2005-09-01

    The Idaho National Engineering Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy is in part due to the history of the INL as the National Reactor Testing Station, in part to its mission to recover highly enriched uranium from SNF and in part to it’s mission to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facility, some dating back 50 years in the site history. The success of the INL SNF program is measured by its ability to: 1) achieve safe existing storage, 2) continue to receive SNF from other locations, both foreign and domestic, 3) repackage SNF from wet storage to interim dry storage, and 4) prepare the SNF for dispositioning in a federal repository. Because of the diversity in the SNF and the facilities at the INL, the INL is addressing almost very condition that may exist in the SNF world. Many of solutions developed by the INL are applicable to other SNF storage sites as they develop their management strategy. The SNF being managed by the INL are in a variety of conditions, from intact assemblies to individual rods or plates to powders, rubble, and metallurgical mounts. Some of the fuel has been in wet storage for over forty years. The fuel is stored bare, or in metal cans and either wet under water or dry in vaults, caissons or casks. Inspections have shown varying degrees of corrosion and degradation of the fuel and the storage cans. Some of the fuel has been recanned under water, and the conditions of the fuel inside the second or third can are unknown. The fuel has been stored in one of 10 different facilities: five wet pools and one casks storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The wet pools range from forty years old to the most modern pool in the US Department of Energy (DOE) complex. The near-term objective is moving the fuel in the older wet storage facilities to

  1. Review of progress in the Canadian nuclear fuel waste management program

    SciTech Connect

    Lyon, R.B.; Johnson, L.H.

    1986-01-01

    The Canadian Nuclear Fuel Waste Management Program is reviewed, illustrating the progress that has been made in assessing the concept of disposal of nuclear fuel waste in plutonic rock of the Canadian Shield. Research is being conducted into used fuel storage and transportation, fuel waste immobilization, site characterization and selection methods, and performance assessment modelling. Details of achievements in these areas are outlined, and results of the most recent interim assessment are discussed.

  2. Intended validation in the Swedish program for spent nuclear fuel

    SciTech Connect

    Cronhjort, B.; Sheng, G.

    1996-12-01

    In September 1992, for the third time, the Swedish Nuclear Fuel and Waste Management Company (SKB) presented its triennial R&D program, as stipulated by the Swedish Act on Nuclear Activities with amendments, e.g. SFS 1995:875. The program, entitled Programme for Research, Development, Demonstration and Other Measures, was carefully reviewed by some forty various national organizations, and universities, as well as by national authorities. In December 1993, the Government decided amongst other things, to ask SKB to develop a validation strategy for models related to the safety assessment of the intended disposal system. Thus, SKB has supplemented its RD&D-Programme 92, and the program with supplement has been reviewed anew by the Swedish Nuclear Power Inspectorate (SKI), and by the National Council for Nuclear Waste (KASAM). SKB has commented upon the reviews. The final Government decision became available on May 18, 1995. In making no comments on SKB`s offered strategy, the Government would seem to reflect tacit approval. The purpose of this paper is to summarize the intended validation strategy. The deliberations related to SKB`s developing the strategy were presented by the authors in a companion paper. In conclusion, the authors relate SKB`s intended strategy to some recent independent research on validation of general system models.

  3. A Practical Approach to a Closed Nuclear Fuel Cycle and Sustained Nuclear Energy - 12383

    SciTech Connect

    Collins, Emory D.; Del Cul, Guillermo D.; Spencer, Barry B.; Williams, Kent A.

    2012-07-01

    Recent systems analysis studies at Oak Ridge National Laboratory (ORNL) have shown that sufficient information is available from previous research and development (R and D), industrial experience, and current studies to make rational decisions on a practical approach to a closed nuclear fuel cycle in the United States. These studies show that a near-term decision is needed to recycle used nuclear fuel (UNF) in the United States, to encourage public recognition that a practical solution to disposal of nuclear energy wastes, primarily UNF, is achievable, and to ensure a focus on essential near-term actions and future R and D. Recognition of the importance of time factors is essential, including the multi-decade time period required to implement industrial-scale fuel recycle at the capacity needed, and the effects of radioactive decay on proliferation resistance, recycling complexity, radioactive emissions, and high-level-waste storage, disposal form development, and eventual emplacement in a geologic repository. Analysis of time factors led to identification of the benefits of processing older fuel and an 'optimum decay storage time'. Further benefits of focused R and D can ensure more complete recycling of UNF components and minimize wastes requiring disposal. Analysis of recycling costs and nonproliferation requirements, which are often cited as reasons for delaying a decision to recycle, shows that (1) the differences in costs of nuclear energy with open or closed fuel cycles are insignificant and (2) nonproliferation requirements can be met by a combination of 'safeguards-by-design' co-location of back-end fuel cycle facilities, and applied engineered safeguards and monitoring. The study shows why different methods of separating and recycling used fuel components do not have a significant effect on nonproliferation requirements and can be selected on other bases, such as process efficiency, maturity, and cost-effectiveness. Finally, the study concludes that

  4. Comparative Study of ADS and FR in Advanced Nuclear Fuel Cycles

    SciTech Connect

    Wydler, Peter; Van Den Durpel, Luc

    2002-07-01

    Several nuclear fuel cycle options may be envisaged in the long term to further improve the sustainability of nuclear energy. Fully closed fuel cycles for actinides can reduce the long-term radiotoxicity of the high-level waste by a hundred-fold compared with current once-through fuel cycles. The OECD/NEA completed a nuclear energy systems study in order to compare the role of accelerator-driven systems and fast reactors in such closed fuel cycles and this with respect to reactor properties, fuel cycle requirements, economic aspects and R and D-needs. (authors)

  5. Applications of Nuclear Data Covariances to Criticality Safety and Spent Fuel Characterization

    NASA Astrophysics Data System (ADS)

    Williams, M. L.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2014-04-01

    Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

  6. Applications of nuclear data covariances to criticality safety and spent fuel characterization

    SciTech Connect

    Williams, Mark L; Ilas, Germina; Marshall, William BJ J; Rearden, Bradley T

    2014-01-01

    Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

  7. Sun, Ocean, Nuclear Bombs, and Fossil Fuels: Radiocarbon Variations and Implications for High-Resolution Dating

    NASA Astrophysics Data System (ADS)

    Dutta, Koushik

    2016-06-01

    Radiocarbon, or 14C, is a radiometric dating method ideally suited for providing a chronological framework in archaeology and geosciences for timescales spanning the last 50,000 years. 14C is easily detectable in most common natural organic materials and has a half-life (5,730±40 years) relevant to these timescales. 14C produced from large-scale detonations of nuclear bombs between the 1950s and the early 1960s can be used for dating modern organic materials formed after the 1950s. Often these studies demand high-resolution chronology to resolve ages within a few decades to less than a few years. Despite developments in modern, high-precision 14C analytical methods, the applicability of 14C in high-resolution chronology is limited by short-term variations in atmospheric 14C in the past. This article reviews the roles of the principal natural drivers (e.g., solar magnetic activity and ocean circulation) and the anthropogenic perturbations (e.g., fossil fuel CO2 and 14C from nuclear and thermonuclear bombs) that are responsible for short-term 14C variations in the environment. Methods and challenges of high-resolution 14C dating are discussed.

  8. Driver Education Saves Gas.

    ERIC Educational Resources Information Center

    American Automobile Association, Falls Church, VA. Traffic Engineering and Safety Dept.

    The argument that driver education should be dropped because driver education cars use gas is shortsighted. High school driver education is an excellent vehicle for teaching concepts of energy conservation. A small investment in fuel now can result in major savings of gasoline over a student's lifetime. In addition good driver education courses…

  9. Apparatus for feeding nuclear fuel pellets to a loading tray

    SciTech Connect

    Huggins, T.B.

    1981-12-08

    Apparatus for feeding nuclear fuel pellets at a uniform, predetermined rate between pellet centering and grinding apparatus and a tray used for loading pellets into a nuclear fuel rod are described. Pellets discharged from the grinder are conveyed by a woven wire belt to a drive wheel which develops a force available to be applied to pellets preceding it on the belt. The pellets pass under the drive wheel which adds additional weight acting vertically on each pellet. This total weight of pellet and drive wheel coupled with wire belt linear movement acts to push a line of about 36 pellets onto a pellet dumping mechanism. As the dumping mechanism is actuated to dump the pellets on to a loading tray, the pellets moving toward the mechanism are clamped in a stationary position and the drive wheel simultaneously is lifted from its pellet contacting position until the pellet dumping process is completed. The clamping device is then lifted from its pellet and the drive wheel simultaneously is lowered into a pellet contacting position.

  10. Microbial studies in the Canadian nuclear fuel waste management program.

    PubMed

    Stroes-Gascoyne, S; West, J M

    1997-07-01

    Atomic Energy of Canada Limited (AECL) has developed a concept for permanent geological disposal of used nuclear fuel in Canada. This concept, based on a multibarrier system, would involve disposal of nuclear fuel waste in titanium or copper containers, surrounded by compacted clay-based buffer and backfill materials, in a vault 500-1000 m deep in granitic rock of the Canadian Shield. Subsurface environments will not be sterile and an experimental program was initiated in 1991 by AECL to address and quantify the potential effects of microbial action on the integrity of the disposal vault. This microbial program focuses on answering specific questions in areas such as the survival of bacteria in compacted clay-based buffer materials under relevant radiation, temperature and desiccation conditions; mobility of microbes in compacted buffer materials; the potential for microbially influenced corrosion of containers; microbial gas production in backfill material; introduction of nutrients as a result of vault excavation and operation; the presence and activity of microbes in deep granitic groundwaters; and the effects of biofilms on radionuclide migration in the geosphere. This paper summarizes the results to date from the research activities at AECL.

  11. Nuclear fuel pellet sintering boat unloading apparatus and method

    SciTech Connect

    Huggins, T.B.; Widener, W.H.; Klapper, K.K.

    1990-05-22

    This patent describes a method for unloading nuclear fuel pellets from a sintering boat having an open top. It comprises: pivoting a transfer housing loaded with the boat filled with nuclear fuel pellets about a generally horizontal axis from an upright position remote from a pellet deposit surface to an inverted position adjacent to the deposit surface to move the boat from an upright to inverted orientation with the pellets retained within the boat by a latched lid in a closed condition on the housing; unlatching the lid of the housing as the housing reaches its inverted position but engaging the unlatched lid with the deposit surface to retain it in its closed condition; and reverse pivoting the housing from its inverted position back toward its upright position to permit the unlatched lid to pivot from the closed condition to an opened condition thereby allowing pellets to slide out of the open top of the inverted boat and down the opened lid of the housing to the deposit site.

  12. Applying fast calorimetry on a spent nuclear fuel calorimeter

    SciTech Connect

    Liljenfeldt, Henrik

    2015-04-15

    Recently at Los Alamos National Laboratory, sophisticated prediction algorithms have been considered for the use of calorimetry for treaty verification. These algorithms aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy. The algorithms have been implemented in MATLAB and applied on existing equilibrium measurements from a spent nuclear fuel calorimeter located at the Swedish nuclear fuel interim storage facility. The results show significant improvements in measurement time in the order of 15 to 50 compared to equilibrium measurements, but cannot predict the heat accurately in less time than the currently used temperature increase method can. This Is both due to uncertainties in the calibration of the method as well as identified design features of the calorimeter that limits the usefulness of equilibrium type measurements. The conclusions of these findings are discussed, and suggestions of both improvements of the current calorimeter as well as what to keep in mind in a new design are given.

  13. Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

    SciTech Connect

    Heard, F.J.; Cramer, E.R.; Beaver, T.R.; Thurgood, M.J.

    1996-01-01

    A series of scoping analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The SNFP was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy`s Hanford Site in Richland, Washington. The subject efforts focused on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms for each of the IPS operations and configurations, obtaining preliminary results for comparison with and verification of other analyses, and providing technology-based recommendations for consideration and incorporation into the design bases for the SNFP. The goal was to develop a series fo thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the SNFP. A series of sensitivity analyses were also performed to help identify those parameters that have the greatest impact on energy transfer and hence, temperature control. It is anticipated that the subject thermal-hydraulic models will form the basis for a series of advanced and more detailed models that will more accurately reflect the thermal performance of the IPS and alleviate the necessity for some of the more conservative assumptions and oversimplifications, as well as form the basis for the final process and safety analyses.

  14. Spent nuclear fuel recycling with plasma reduction and etching

    DOEpatents

    Kim, Yong Ho

    2012-06-05

    A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF4 or NF3 as the fluorine sources instead of F2 or HF.

  15. Behavior of iodine in the dissolution of spent nuclear fuels

    SciTech Connect

    Sakurai, Tsutomu; Komatsu, Kazunori; Takahashi, A.

    1997-08-01

    The results of laboratory-scale experiments concerning the behavior of iodine in the dissolution of spent nuclear fuels, which were carried out at the Japan Atomic Energy Research Institute, are summarized. Based on previous and new experimental results, the difference in quantity of residual iodine in the fuel solution between laboratory-scale experiments and reprocessing plants is discussed, Iodine in spent fuels is converted to the following four states: (1) oxidation into I{sub 2} by nitric acid, (2) oxidation into I{sub 2} by nitrous acid generated in the dissolution, (3) formation of a colloid of insoluble iodides such as AgI and PdI{sub 2}, and (4) deposition on insoluble residue. Nitrous acid controls the amount of colloid formed. As a result, up to 10% of iodine in spent fuels is retained in the fuel solution, up to 3% is deposited on insoluble residue, and the balance volatilizes to the off-gas, Contrary to earlier belief, when the dissolution is carried out in 3 to 4 M HNO{sub 3} at 100{degrees}C, the main iodine species in a fuel solution is a colloid, not iodate, Immediately after its formation, the colloid is unstable and decomposes partially in the hot nitric acid solution through the following reaction: AgI(s) + 2HNO{sub 3}(aq) = {1/2}I{sub 2}(aq) + AgNO{sub 3}(aq) + NO{sub 2}(g) + H{sub 2}O(1). For high concentrations of gaseous iodine, I{sub 2}(g), and NO{sub 2}, this reaction is reversed towards formation of the colloid (AgI). Since these concentrations are high near the liquid surface of a plant-scale dissolver, there is a possibility that the colloid is formed there through this reversal, Simulations performed in laboratory-scale experiments demonstrated this reversal, This phenomenon can be one reason the quantity of residual iodine in spent fuels is higher in reprocessing plants than in laboratory-scale experiments. 17 refs., 5 figs., 3 tabs.

  16. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    SciTech Connect

    Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.; Massaro, Lawrence M.; Jensen, Philip J.

    2014-10-01

    This report presents a preliminary evaluation of removing used nuclear fuel (UNF) from 12 shutdown nuclear power plant sites. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. The evaluation was divided into four components: characterization of the UNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory; a description of the on-site infrastructure and conditions relevant to transportation of UNF and GTCC waste; an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing UNF and GTCC waste, including identification of gaps in information; and, an evaluation of the actions necessary to prepare for and remove UNF and GTCC waste. The primary sources for the inventory of UNF and GTCC waste are the U.S. Department of Energy (DOE) RW-859 used nuclear fuel inventory database, industry sources such as StoreFUEL and SpentFUEL, and government sources such as the U.S. Nuclear Regulatory Commission. The primary sources for information on the conditions of site and near-site transportation infrastructure and experience included observations and information collected during visits to the Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion sites; information provided by managers at the shutdown sites; Facility Interface Data Sheets compiled for DOE in 2005; Services Planning Documents prepared for DOE in 1993 and 1994; industry publications such as Radwaste Solutions; and Google Earth. State and Regional Group representatives, a Tribal representative, and a Federal Railroad Administration representative participated in six of the shutdown site

  17. Used nuclear fuel separations process simulation and testing

    SciTech Connect

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D.

    2013-07-01

    Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

  18. REACTOR PHYSICS MODELING OF SPENT NUCLEAR RESEARCH REACTOR FUEL FOR SNM ATTRIBUTION AND NUCLEAR FORENSICS

    SciTech Connect

    Sternat, M.; Beals, D.; Webb, R.; Nichols, T.

    2010-06-09

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle. The

  19. 77 FR 59994 - Nuclear Fuel Services, Inc., Erwin, TN; Issuance of License Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-01

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Nuclear Fuel Services, Inc., Erwin, TN; Issuance of License Renewal AGENCY: U.S. Nuclear... M. Ramsey, Project Manager, Office of Nuclear Material Safety and Safeguards, U.S....

  20. Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

    SciTech Connect

    Trammell, Michael P; Jolly, Brian C; Miller, James Henry; Qualls, A L; Harrison, Thomas J

    2013-01-01

    ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

  1. Managing Spent Nuclear Fuel at the Idaho National Laboratory

    SciTech Connect

    Thomas Hill; Denzel L. Fillmore

    2005-10-01

    The Idaho National Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy derives from the history of the INL as the National Reactor Testing Station, and from its mission to recover HEU from SNF and to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facilities, some 50 years old. SNF at INL has many forms—from intact assemblies down to metallurgical mounts, and some fuel has been wet stored for over 40 years. SNF is stored bare or in metal cans under water, or dry in vaults, caissons or casks. Inspection shows varying corrosion and degradation of the SNF and its storage cans. SNF has been stored in 10 different facilities: 5 pools, one cask storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The pools range in age from 40 years old to the most modern in the US Department of Energy (DOE) complex. The near-term objective is to move SNF from older pools to interim dry storage, allowing shutdown and decommissioning of the older facilities. This move involves drying methods that are dependent on fuel type. The long-term objective is to have INL SNF in safe dry storage and ready to be shipped to the National Repository. The unique features of the INL SNF requires special treatments and packaging to meet the proposed repository acceptance criteria and SNF will be repackaged in standardized canisters for shipment and disposal in the National Repository. Disposal will use the standardized canisters that can be co-disposed with High Level Waste glass logs to limit the total fissile material in a repository waste package. The DOE standardized canister also simplifies the repository handling of the multitude of DOE SNF sizes and shapes.

  2. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    SciTech Connect

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

  3. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    SciTech Connect

    McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  4. Implementation process and deployment initiatives for the regionalized storage of DOE-owned spent nuclear fuel

    SciTech Connect

    Dearien, J.A.; Smith, N.E.L.

    1995-12-31

    This report describes how DOE-owned spent nuclear fuel (SNF) will be stored in the interim 40-year period from 1996 to 2035, by which time it is expected to be in a National Nuclear Repository. The process is described in terms of its primary components: fuel inventory, facilities where it is stored, how the fuel will be moved, and legal issues associated with the process. Tools developed to deploy and fulfill the implementation needs of the National Spent Nuclear Fuel Program are also discussed.

  5. Behavior of spent nuclear fuel and storage system components in dry interim storage. Revision 1

    SciTech Connect

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1983-02-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom; organic-cooled reactor (OCR) fuel (clad with a zirconium alloy) in silos in Canada; and boiling water reactor (BWR) fuel (clad with Zircaloy) in a metal storage cask in Germany. Dry storage demonstrations are under way for Zircaloy-clad fuel from BWRs, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions. 110 refs., 22 figs., 28 tabs.

  6. High temperature dilatometry of simulated oxide nuclear fuel

    NASA Astrophysics Data System (ADS)

    Tenishev, A. V.; Baranov, V. G.; Kuzmin, R. S.; Pokrovskiy, S. A.

    2016-04-01

    High temperature dilatometry of model systems based on uranium dioxide with additives of burnable neutron absorbers both as Gd2O3 and as AlGdO3, and fission products simulators (FPS) was performed. It shown that in some cases instead of high temperature samples shrinkage there is a sharp transition to the expansion, which is associated with an increase of the samples volume due to the formation of liquid phases. The beginning of a complex composition eutectic melting starts at temperatures from 1950 to 2250 °C in the uranium dioxide samples containing significant amounts of Al, Gd, and FPS. Thus, in the analysis of oxide nuclear fuel behavior at high temperatures should be considered that the formation of liquid phases is possible at a temperature of 1000 °C lower than a melting point of pure stoichiometric uranium dioxide if its initial composition became more complex.

  7. Eddy Current Examination of Spent Nuclear Fuel Canister Closure Welds

    SciTech Connect

    Arthur D. Watkins; Dennis C. Kunerth; Timothy R. McJunkin

    2006-04-01

    The National Spent Nuclear Fuel Program (NSNFP) has developed standardized DOE SNF canisters for handling and interim storage of SNF at various DOE sites as well as SNF transport to and SNF handling and disposal at the repository. The final closure weld of the canister will be produced remotely in a hot cell after loading and must meet American Society of Mechanical Engineers (ASME) Section III, Division 3 code requirements thereby requiring volumetric and surface nondestructive evaluation to verify integrity. This paper discusses the use of eddy current testing (ET) to perform surface examination of the completed welds and repair cavities. Descriptions of integrated remote welding/inspection system and how the equipment is intended function will also be discussed.

  8. Air Transport of Spent Nuclear Fuel (SNF) Assemblies

    SciTech Connect

    Haire, M.J.; Moses, S.D.; Shapovalov, V.I.; Morenko, A.

    2007-07-01

    Sometimes the only feasible means of shipping research reactor spent nuclear fuel (SNF) among countries is via air transport because of location or political conditions. The International Atomic Energy Agency (IAEA) has established a regulatory framework to certify air transport Type C casks. However, no such cask has been designed, built, tested, and certified. In lieu of an air transport cask, research reactor SNF has been transported using a Type B cask under an exemption with special arrangements for administrative and security controls. This work indicates that it may be feasible to transport commercial power reactor SNF assemblies via air, and that the cost is only about three times that of shipping it by railway. Optimization (i.e., reduction) of this cost factor has yet to be done. (authors)

  9. Investigation of Electrochemical Recovery of Zirconium from Spent Nuclear Fuels

    SciTech Connect

    Simpson, Michael; Hwang, II-Soon

    2014-06-01

    This project uses both modeling and experimental studies to design optimal electrochemical technology methods for recovery of zirconium from used nuclear fuel rods for more effective waste management. The objectives are to provide a means of efficiently separating zirconium into metallic high-level waste forms and to support development of a process for decontamination of zircaloy hulls to enable their disposal as low- and intermediate-level waste. Modeling work includes extension of a 3D model previously developed by Seoul National University for uranium electrorefining by adding the ability to predict zirconium behavior. Experimental validation activities include tests for recovery of zirconium from molten salt solutions and aqueous tests using surrogate materials. *This is a summary of the FY 2013 progress for I-NERI project # 2010-001-K provided to the I-NERI office.

  10. MICROBIAL TRANSFORMATIONS OF RADIONUCLIDES RELEASED FROM NUCLEAR FUEL REPROCESSING PLANTS.

    SciTech Connect

    FRANCIS,A.J.

    2006-10-18

    Microorganisms can affect the stability and mobility of the actinides U, Pu, Cm, Am, Np, and the fission products Tc, I, Cs, Sr, released from nuclear fuel reprocessing plants. Under appropriate conditions, microorganisms can alter the chemical speciation, solubility and sorption properties and thus could increase or decrease the concentrations of radionuclides in solution and the bioavailability. Dissolution or immobilization of radionuclides is brought about by direct enzymatic action or indirect non-enzymatic action of microorganisms. Although the physical, chemical, and geochemical processes affecting dissolution, precipitation, and mobilization of radionuclides have been investigated, we have only limited information on the effects of microbial processes. The mechanisms of microbial transformations of the major and minor actinides and the fission products under aerobic and anaerobic conditions in the presence of electron donors and acceptors are reviewed.

  11. Fuel areal density distributions derived from nuclear scattering signatures

    NASA Astrophysics Data System (ADS)

    Bionta, R. M.; Casey, D. T.; Cerjan, C. J.; Yeamans, C. B.; Gatu Johnson, M. G.

    2016-10-01

    The spatial variation of activities measured in the array of 20 Nuclear Activation Detectors mounted on the flanges around the NIF target chamber (FNADs) are correlated with asymmetries in the underlying fuel areal density of compressed ICF targets. The asymmetric areal density distributions cause variations in the neutron spectra with direction which are seen in the dsr (down scattered ratio) metric, the ratio of the number of 10-12 MeV neutrons to the number of 13-15 MeV neutrons. We show, using a simple physics based simulation of neutron scattering through an idealized non-uniform DT shell with a realistic neutron source, that for most shots an areal distribution can be found which reproduces both the FNAD activity and the dsr measurements. Furthermore, by linking the simulation to a Marquardt minimizer, we fit the areal distribution to a truncated set of spherical harmonics. Prepared by LLNL under Contract DE-AC52-07NA27344.

  12. National briefing summaries: Nuclear fuel cycle and waste management

    SciTech Connect

    Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

    1988-12-01

    The National Briefing Summaries is a compilation of publicly available information concerning the nuclear fuel cycle and radioactive waste management strategies and programs of 21 nations, including the United States and three international agencies that have publicized their activities in this field. It presents available highlight information with references that may be used by the reader for additional information. The information in this document is compiled primarily for use by the US Department of Energy and other US federal agencies and their contractors to provide summary information on radioactive waste management activities in other countries. This document provides an awareness to managers and technical staff of what is occurring in other countries with regard to strategies, activities, and facilities. The information may be useful in program planning to improve and benefit United States' programs through foreign information exchange. Benefits to foreign exchange may be derived through a number of exchange activities.

  13. The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    NASA Astrophysics Data System (ADS)

    Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

    2017-01-01

    The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

  14. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Jiang, Hao; Bevard, Bruce Balkcom; Howard, Rob L; Scaglione, John M

    2015-01-01

    The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into local stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

  15. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    SciTech Connect

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  16. Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities

    SciTech Connect

    Soelberg, Nick R.; Garn, Troy G.; Greenhalgh, Mitchell R.; Law, Jack D.; Jubin, Robert; Strachan, Denis M.; Thallapally, Praveen K.

    2013-01-01

    The removal of volatile radionuclides generated during used nuclear fuel reprocessing in the US is almost certain to be necessary for the licensing of a reprocessing facility in the US. Various control technologies have been developed, tested, or used over the past 50 years for control of volatile radionuclide emissions from used fuel reprocessing plants. The US DOE has sponsored, since 2009, an Off-gas Sigma Team to perform research and development focused on the most pressing volatile radionuclide control and immobilization problems. In this paper, we focus on the control requirements and methodologies for85Kr and129I. Numerous candidate technologies have been studied and developed at laboratory and pilot-plant scales in an effort to meet the need for high iodine control efficiency and to advance alternatives to cryogenic separations for krypton control. Several of these show promising results. Iodine decontamination factors as high as 105, iodine loading capacities, and other adsorption parameters including adsorption rates have been demonstrated under some conditions for both silver zeolite (AgZ) and Ag-functionalized aerogel. Sorbents, including an engineered form of AgZ and selected metal organic framework materials (MOFs), have been successfully demonstrated to capture Kr and Xe without the need for separations at cryogenic temperatures.

  17. Dose reduction improvements in storage basins of spent nuclear fuel

    SciTech Connect

    Huang, Fan-Hsiung F.

    1997-08-13

    Spent nuclear fuel in storage basins at the Hanford Site has corroded and contaminated basin water, which has leaked into the soil; the fuel also had deposited a layer of radioactive sludge on basin floors. The SNF is to be removed from the basins to protect the nearby Columbia River. Because the radiation level is high, measures have been taken to reduce the background dose rate to as low as reasonably achievable (ALARA) to prevent radiation doses from becoming the limiting factor for removal of the SW in the basins to long-term dry storage. All activities of the SNF Project require application of ALARA principles for the workers. On the basis of these principles dose reduction improvements have been made by first identifying radiological sources. Principal radiological sources in the basin are basin walls, basin water, recirculation piping and equipment. Dose reduction activities focus on cleaning and coating basin walls to permit raising the water level, hydrolasing piping, and placing lead plates. In addition, the transfer bay floor will be refinished to make decontamination easier and reduce worker exposures in the radiation field. The background dose rates in the basin will be estimated before each task commences and after it is completed; these dose reduction data will provide the basis for cost benefit analysis.

  18. Increase of inherent protection level in spent nuclear fuel

    SciTech Connect

    Krasnobaev, A.; Kryuchkov, E.; Glebov, V.

    2006-07-01

    The paper is devoted to upgrading inherent proliferation protection of fissionable nuclear materials (FNM). Some possibilities were investigated to form high radiation barrier inside spent fuel assemblies (SFA) discharged from power reactors of VVER-1000 type and research reactors of IRT type. The radiation barrier is estimated in the terms of rate of equivalent dose (RED) at 30-cm distance from SFA. The values of RED were calculated with application of the computer code package SCALE 4.3. The paper considers the criteria adopted for estimation of FNM proliferation resistance. The paper presents numerical results on a component-wise analysis of the radiation barrier in SFA from reactors of VVER-1000 and IRT type and on capability of various radionuclides to prolong action of the radiation barrier. Isotopic admixtures were selected and amounts of these admixtures were evaluated for significant prolongation of the radiation barrier action at the levels of the radiation standards used for estimation of FNM proliferation resistance. The paper considers vulnerability of the radiation barriers in respect to thermal processing of spent fuel. (authors)

  19. Software Design Document for the AMP Nuclear Fuel Performance Code

    SciTech Connect

    Philip, Bobby; Clarno, Kevin T; Cochran, Bill

    2010-03-01

    The purpose of this document is to describe the design of the AMP nuclear fuel performance code. It provides an overview of the decomposition into separable components, an overview of what those components will do, and the strategic basis for the design. The primary components of a computational physics code include a user interface, physics packages, material properties, mathematics solvers, and computational infrastructure. Some capability from established off-the-shelf (OTS) packages will be leveraged in the development of AMP, but the primary physics components will be entirely new. The material properties required by these physics operators include many highly non-linear properties, which will be replicated from FRAPCON and LIFE where applicable, as well as some computationally-intensive operations, such as gap conductance, which depends upon the plenum pressure. Because there is extensive capability in off-the-shelf leadership class computational solvers, AMP will leverage the Trilinos, PETSc, and SUNDIALS packages. The computational infrastructure includes a build system, mesh database, and other building blocks of a computational physics package. The user interface will be developed through a collaborative effort with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Capability Transfer program element as much as possible and will be discussed in detail in a future document.

  20. Commercial Nuclear Fuel Leasing - The Relationships to Nonproliferation and Repository Site Performance

    SciTech Connect

    Pentz, D.L.; Stoll, R.H.

    2007-07-01

    This paper describes the authors' concept of nuclear fuel leasing - - commercially-based and market-driven - - for nuclear power plant (NPP) facilities. Key issues currently affecting further development of this fuel leasing concept are examined, including issues of nonproliferation contribution and spent fuel management. If the nuclear power renaissance is to be realized in conjunction with a serious effort to reduce the impacts of greenhouse gas emissions from increasing electricity demand, nuclear fuel leasing is an important option for the current fuel cycle by its ability to extend the positive benefits of the current nonproliferation regime to countries where the scale of small programs and the complexities of the geology make final disposition so challenging. The authors believe that a principal focus on commercial options for nuclear fuel leasing is essential in order to make these options sustainable and acceptable, especially in countries wanting to build nuclear power plants to meet energy demands in an internationally acceptable way and to meet international concerns for preventing further proliferation of nuclear weapons technology and for reducing climate change effects. The authors are unaware of any public documents that describe market-priced and commercially-driven examples for fuel leasing. This paper discusses the main elements and issues for commercial fuel leasing based on detailed examinations of several conceptual models during the past eight years. (authors)

  1. Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

    NASA Astrophysics Data System (ADS)

    Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

    2005-02-01

    A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

  2. Behavior of spent nuclear fuel and storage system components in dry interim storage.

    SciTech Connect

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1982-08-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions.

  3. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    SciTech Connect

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  4. International Source Book: Nuclear Fuel Cycle Research and Development Vol 1 Volume 1

    SciTech Connect

    Harmon, K. M.; Lakey, L. T.

    1983-07-01

    This document starts with an overview that summarizes nuclear power policies and waste management activities for nations with significant commercial nuclear fuel cycle activities either under way or planned. A more detailed program summary is then included for each country or international agency conducting nuclear fuel cycle and waste management research and development. This first volume includes the overview and the program summaries of those countries listed alphabetically from Argentina to Italy.

  5. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    SciTech Connect

    Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.; Massaro, Lawrence M.; Jensen, Philip J.

    2015-09-30

    A preliminary evaluation of removing spent nuclear fuel (SNF) from 13 shutdown nuclear power reactor sites was conducted. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, and Vermont Yankee. The evaluation was divided into four components: (1) characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory, (2) a description of the on-site infrastructure and conditions relevant to transportation of SNF and GTCC waste, (3) an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing SNF and GTCC waste, including identification of gaps in information, and (4) an evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. Every site was found to have at least one off-site transportation mode option for removing its SNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

  6. Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options.

    SciTech Connect

    Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J.; Parma, Edward J., Jr.; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

    2007-10-01

    The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents.

  7. Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities

    SciTech Connect

    Soelberg, Nicolas R.; Garn, Troy; Greenhalgh, Mitchell; Law, Jack; Jubin, Robert T.; Strachan, Denis M.; Thallapally, Praveen K.

    2013-07-22

    Nuclear fission results in the production of fission products and activation products, some of which tend to be volatile during used fuel reprocessing. These can evolve in volatile species in the reprocessing facility off-gas streams, depending on the separations and reprocessing technologies that are used. Radionuclides that have been identified as “volatile radionuclides” are noble gases (most notably isotopes of Kr and Xe); 3H; 14C; and 129I. Radionuclides that tend to form volatile species that evolve into reprocessing facility off-gas systems are more challenging to efficiently control compared to radionuclides that tend to stay in solid or liquid phases. Future used fuel reprocessing facilities in the United States can require efficient capture of some volatile radionuclides in their off-gas streams to meet regulatory emission requirements. In aqueous reprocessing, these radionuclides are most commonly expected to evolve into off-gas streams in tritiated water [3H2O (T2O) and 3HHO (THO)], radioactive CO2, noble gases, and gaseous HI, I2, or volatile organic iodides. The fate and speciation of these radionuclides from a non-aqueous fuel reprocessing facility is less well known at this time, but active investigations are in progress. An Off-Gas Sigma Team was formed in late FY 2009 to integrate and coordinate the Fuel Cycle Research and Development (FCR&D) activities directed towards the capture and sequestration of the these volatile radionuclides (Jubin 2012a). The Sigma Team concept was envisioned to bring together multidisciplinary teams from across the DOE complex that would work collaboratively to solve the technical challenges and to develop the scientific basis for the capture and immobilization technologies such that the sum of the efforts was greater than the individual parts. The Laboratories currently participating in this effort are Argonne National Laboratory (ANL), Idaho National Laboratory (INL), Oak Ridge National Laboratory (ORNL), Pacific

  8. The Nuclear Fuel Services, Inc. program to support disposition of enriched uranium-bearing materials

    SciTech Connect

    Schutt, Stephen M.; Jacob, Norman P.

    2007-07-01

    The disposition of surplus nuclear materials has become one of the most pressing issues of our time. Numerous agencies have invoked programs with the purpose of removing such materials from various international venues and disposing these materials in a manner that achieves non-proliferability. This paper describes the Nuclear Fuel Services, Inc (NFS) Nuclear Material Disposition Program, which to date has focused on a variety of Special Nuclear Material (SNM), in particular uranium of various enrichments. The major components of this program are discussed, with emphasis on recycle and return of material to the nuclear fuel cycle. (authors)

  9. System analyses on advanced nuclear fuel cycle and waste management

    NASA Astrophysics Data System (ADS)

    Cheon, Myeongguk

    To evaluate the impacts of accelerator-driven transmutation of waste (ATW) fuel cycle on a geological repository, two mathematical models are developed: a reactor system analysis model and a high-level waste (HLW) conditioning model. With the former, fission products and residual trans-uranium (TRU) contained in HLW generated from a reference ATW plant operations are quantified and the reduction of TRU inventory included in commercial spent-nuclear fuel (CSNF) is evaluated. With the latter, an optimized waste loading and composition in solidification of HLW are determined and the volume reduction of waste packages associated with CSNF is evaluated. WACOM, a reactor system analysis code developed in this study for burnup calculation, is validated by ORIGEN2.1 and MCNP. WACOM is used to perform multicycle analysis for the reference lead-bismuth eutectic (LBE) cooled transmuter. By applying the results of this analysis to the reference ATW deployment scenario considered in the ATW roadmap, the HLW generated from the ATW fuel cycle is quantified and the reduction of TRU inventory contained in CSNF is evaluated. A linear programming (LP) model has been developed for determination of an optimized waste loading and composition in solidification of HLW. The model has been applied to a US-defense HLW. The optimum waste loading evaluated by the LP model was compared with that estimated by the Defense Waste Processing Facility (DWPF) in the US and a good agreement was observed. The LP model was then applied to the volume reduction of waste packages associated with CSNF. Based on the obtained reduction factors, the expansion of Yucca Mountain Repository (YMR) capacity is evaluated. It is found that with the reference ATW system, the TRU contained in CSNF could be reduced by a factor of ˜170 in terms of inventory and by a factor of ˜40 in terms of toxicity under the assumed scenario. The number of waste packages related to CSNF could be reduced by a factor of ˜8 in terms of

  10. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    DOEpatents

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  11. Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

    NASA Technical Reports Server (NTRS)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar

    2012-01-01

    A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

  12. The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication

    SciTech Connect

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

    2009-02-01

    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.

  13. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    SciTech Connect

    Adrian Miron; Joshua Valentine; John Christenson; Majd Hawwari; Santosh Bhatt; Mary Lou Dunzik-Gougar: Michael Lineberry

    2009-10-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  14. Determination of nuclear fuel burn-up axial profile by neutron emission measurement

    NASA Astrophysics Data System (ADS)

    Prokopowicz, Rafal; Pytel, Krzysztof

    2016-12-01

    Burning-up of nuclear fuel is usually not a space-isotropic phenomenon. It depends on both the neutron flux density and energy spectrum distribution during fuel operation in a nuclear reactor. This paper presents the method of measurement of burn-up spatial distribution of spent nuclear fuel element. The method is based on recording of the neutron emission from investigated fuel element. Based on performed analyses and calculations, a suitable measuring setup has been designed and constructed. The subjects of investigation were fuel elements used in the MARIA research reactor, operated by National Centre for Nuclear Research in Świerk, Poland. The results of measurements made over a period of several years by means of the described method are presented in the paper.

  15. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    SciTech Connect

    Clarno, Kevin T; Hamilton, Steven P; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth; Pugmire, Dave; Dilts, Gary; Banfield, James E

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

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

    SciTech Connect

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

    2009-03-01

    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.

  17. Direct Investigations of the Immobilization of Radionuclides in the Alteration Products of Spent Nuclear Fuel

    SciTech Connect

    Peter C. Burns; Robert J. Finch; David J. Wronkiewicz

    2004-12-27

    Safe disposal of the nation's nuclear waste in a geological repository involves unique scientific and engineering challenges owing to the very long-lived radioactivity of the waste. The repository must retain a variety of radionuclides that have vastly different chemical characters for several thousand years. Most of the radioactivity that will be housed in the proposed repository at Yucca Mountain will be associated with spent nuclear fuel, much of which is derived from commercial reactors. DOE is custodian of approximately 8000 tons of spent nuclear fuel that is also intended for eventual disposal in a geological repository. Unlike the spent fuel from commercial reactors, the DOE fuel is diverse in composition with more than 250 varieties. Safe disposal of spent fuel requires a detailed knowledge of its long-term behavior under repository conditions, as well as the fate of radionuclides released from the spent fuel as waste containers are breached.

  18. Nuclear Fuel Cycle Options Evaluation to Inform R&D Planning

    SciTech Connect

    R. Wigeland; T. Taiwo; M. Todosow; H. Ludewig; W. Halsey; J. Gehin; R. Jubin; J. Buelt; S. Stockinger; K. Jenni; B. Oakley

    2014-04-01

    An Evaluation and Screening (E&S) of nuclear fuel cycle options has been conducted in fulfilment of a Charter specified for the study by the U.S. Department of Energy (DOE) Office of Nuclear Energy. The E&S study used an objective and independently reviewed evaluation process to provide information about the potential benefits and challenges that could strengthen the basis and provide guidance for the research and development(R&D) activities undertaken by the DOE Fuel Cycle Technologies Program Office. Using the nine evaluation criteria specified in the Charter and associated evaluation metrics and processes developed during the E&S study, a screening was conducted of 40 nuclear fuel cycle evaluation groups to provide answers to the questions: (1) Which nuclear fuel cycle system options have the potential for substantial beneficial improvements in nuclear fuel cycle performance, and what aspects of the options make these improvements possible? (2)Which nuclear material management approaches can favorably impact the performance of fuel cycle options? (3)Where would R&D investment be needed to support the set of promising fuel cycle system options and nuclear material management approaches identified above, and what are the technical objectives of associated technologies?

  19. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    SciTech Connect

    Wang, Jy-An John; Jiang, Hao

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  20. Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

    SciTech Connect

    Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.; Wood, Thomas W.; Perkins, Casey J.

    2011-11-14

    For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may be constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.

  1. Overview of reductants utilized in nuclear fuel reprocessing/recycling

    SciTech Connect

    Patricia Paviet-Hartmann; Catherine Riddle; Keri Campbell; Edward Mausolf

    2013-10-01

    Most of the aqueous processes developed, or under consideration worldwide for the recycling of used nuclear fuel (UNF) utilize the oxido-reduction properties of actinides to separate them from other radionuclides. Generally, after acid dissolution of the UNF, (essentially in nitric acid solution), actinides are separated from the raffinate by liquid-liquid extraction using specific solvents, associated along the process, with a particular reductant that will allow the separation to occur. For example, the industrial PUREX process utilizes hydroxylamine as a plutonium reductant. Hydroxylamine has numerous advantages: not only does it have the proper attributes to reduce Pu(IV) to Pu(III), but it is also a non-metallic chemical that is readily decomposed to innocuous products by heating. However, it has been observed that the presence of high nitric acid concentrations or impurities (such as metal ions) in hydroxylamine solutions increase the likelihood of the initiation of an autocatalytic reaction. Recently there has been some interest in the application of simple hydrophilic hydroxamic ligands such as acetohydroxamic acid (AHA) for the stripping of tetravalent actinides in the UREX process flowsheet. This approach is based on the high coordinating ability of hydroxamic acids with tetravalent actinides (Np and Pu) compared with hexavalent uranium. Thus, the use of AHA offers a route for controlling neptunium and plutonium in the UREX process by complexant based stripping of Np(IV) and Pu(IV) from the TBP solvent phase, while U(VI) ions are not affected by AHA and remain solvated in the TBP phase. In the European GANEX process, AHA is also used to form hydrophilic complexes with actinides and strip them from the organic phase into nitric acid. However, AHA does not decompose completely when treated with nitric acid and hampers nitric acid recycling. In lieu of using AHA in the UREX + process, formohydroxamic acid (FHA), although not commercially available, hold

  2. Nuclear fuel cycle assessment of India: A technical study for U.S.-India cooperation

    NASA Astrophysics Data System (ADS)

    Krishna, Taraknath Woddi Venkat

    The recent civil nuclear cooperation proposed by the Bush Administration and the Government of India has heightened the necessity of assessing India's nuclear fuel cycle inclusive of nuclear materials and facilities. This agreement proposes to change the long-standing U.S. policy of preventing the spread of nuclear weapons by denying nuclear technology transfer to non-NPT signatory states. The nuclear tests in 1998 have convinced the world community that India would never relinquish its nuclear arsenal. This has driven the desire to engage India through civilian nuclear cooperation. The cornerstone of any civilian nuclear technological support necessitates the separation of military and civilian facilities. A complete nuclear fuel cycle assessment of India emphasizes the entwinment of the military and civilian facilities and would aid in moving forward with the separation plan. To estimate the existing uranium reserves in India, a complete historical assessment of ore production, conversion, and processing capabilities was performed using open source information and compared to independent reports. Nuclear energy and plutonium production (reactor- and weapons-grade) was simulated using declared capacity factors and modern simulation tools. The three-stage nuclear power program entities and all the components of civilian and military significance were assembled into a flowsheet to allow for a macroscopic vision of the Indian fuel cycle. A detailed view of the nuclear fuel cycle opens avenues for technological collaboration. The fuel cycle that grows from this study exploits domestic thorium reserves with advanced international technology and optimized for the existing system. To utilize any appreciable fraction of the world's supply of thorium, nuclear breeding is necessary. The two known possibilities for production of more fissionable material in the reactor than is consumed as fuel are fast breeders or thermal breeders. This dissertation analyzes a thermal

  3. Eco-hydrological and climatic drivers of fuel moisture dynamics in complex terrain

    NASA Astrophysics Data System (ADS)

    Nyman, Petter; Duff, Thomas; Baillie, Craig; Sheridan, Gary

    2016-04-01

    Fuel moisture is a critical parameter for predicting fire behaviour and for planning prescribed burning operations. Moisture content in fuels located on or near the forest floor is particularly important because this fuel source 1) can comprise a large component of the overall fuel load, 2) can have a strong impact on fire spread, and 3) it can in many cases be effectively managed with prescribed burning. Being able to predict surface fuel moisture content is therefore an important research topic. Moisture dynamics in surface fuel are a function of microclimate above the litter layer, rainfall, interception, soil moisture and the hydraulic properties of the fuel itself. Process-based fuel moisture models include these factors in their predictions. However, the data needed to parametrise and test such models at landscape scales are often lacking. The relative importance of various components of the water balance in the litter layer is therefore unknown. In this research we seek to quantify how climate, vegetation and eco-hydrological feedback contribute to variation in net radiation and potential evaporation at the forest floor. Research sites were established at 16 locations in eucalyptus forests in south-east Australia with variable elevation, solar exposure, and drainage areas. Forests ranged from open woodland to tall temperate forests. At these sites we measured solar radiation, air temperature, relative humidity, throughfall, litter moisture, soil moisture, and litter temperature. Forest structure was characterised using hemispherical photos. Using these data on microclimate and vegetation structure we develop and parametrise a Penman-Monteith model of potential evaporation on the forest floor at daily timescales. Using this model of potential evaporation alongside landscape-scale information on the long term water and energy balance we quantify the effects of topography, long-term climate and eco-hydrological feedback on the energy and water balance at the

  4. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    SciTech Connect

    Francis, Matthew W.; Weber, Charles F.; Pigni, Marco T.; Gauld, Ian C.

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  5. Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

    DOE PAGES

    Hu, Jianwei; Gauld, Ian C.

    2014-12-01

    The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

  6. Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

    SciTech Connect

    Hu, Jianwei; Gauld, Ian C.

    2014-12-01

    The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried out to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.

  7. INEL integrated spent nuclear fuel consolidation task team report

    SciTech Connect

    Henry, R.N.; Clark, J.H.; Chipman, N.A.

    1994-09-12

    This document describes a draft plan and schedule to consolidate spent nuclear fuel (SNF) and special nuclear material (SNW) from aging storage facilities throughout the Idaho National Engineering Laboratory (INEL) to the Idaho Chemical Processing Plant (ICPP) in a safe, cost-effective, and expedient manner. A fully integrated and resource-loaded schedule was developed to achieve consolidation as soon as possible. All of the INEL SNF and SNM management task, projects, and related activities from fiscal year 1994 to the end of the consolidation period are logic-tied and integrated with each other. The schedule and plan are presented to initiate discussion of their implementation, which is expected to generate alternate concepts that can be evaluated using the methodology described in this report. Three perturbations to consolidating SNF as soon as possible are also explored. If the schedule is executed as proposed, the new and on-going consolidation activities will require about 6 years to complete and about $25.3M of additional funding. Reduced annual operating costs are expected to recover the additional investment in about 6.4 years. The total consolidation program as proposed will cost about $66.8M and require about 6 years to recover via reduced operating costs from retired SNF/SNM storage facilities. Detailed schedules and cost estimates for the Test Reactor Area Materials Test Reactor canal transfers are included as an example of the level of detail that is typical of the entire schedule (see Appendix D). The remaining work packages for each of the INEL SNF consolidation transfers are summarized in this document. Detailed cost and resource information is available upon request for any of the SNF consolidation transfers.

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

    SciTech Connect

    E. R. Johnson; R. E. Best

    2009-12-28

    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

  9. Lateral Flow Field Behavior Downstream of Mixing Vanes In a Simulated Nuclear Fuel Rod Bundle

    SciTech Connect

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2004-07-01

    To assess the fuel assembly performance of PWR nuclear fuel assemblies, average subchannel flow values are used in design analyses. However, for this highly complex flow, it is known that local conditions around fuel rods vary dependent upon the location of the fuel rod in the fuel assembly and upon the support grid design that maintains the fuel rod pitch. To investigate the local flow in a simulated nuclear fuel rod bundle, a testing technique has been employed to measure the lateral flow field in a 5 x 5 rod bundle. Particle Image Velocimetry was used to measure the lateral flow field downstream of a support grid with mixing vanes for four unique subchannels in the 5 x 5 bundle. The dominant lateral flow structures for each subchannel are compared in this paper including the decay of these flow structures. (authors)

  10. On feasibility of a closed nuclear power fuel cycle with minimum radioactivity

    SciTech Connect

    Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F.

    2015-12-15

    Practical implementation of a closed nuclear fuel cycle implies solution of two main tasks. The first task is creation of environmentally acceptable operating conditions of the nuclear fuel cycle considering, first of all, high radioactivity of the involved materials. The second task is creation of effective and economically appropriate conditions of involving fertile isotopes in the fuel cycle. Creation of technologies for management of the high-level radioactivity of spent fuel reliable in terms of radiological protection seems to be the hardest problem.

  11. ORNL experience and perspectives related to processing of thorium and 233U for nuclear fuel

    DOE PAGES

    Croff, Allen G.; Collins, Emory D.; Del Cul, G. D.; ...

    2016-05-01

    Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233U for recycle in thermal reactors, fast reactors, or externally driven systems. Here, lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233U.

  12. Status of radioiodine control for nuclear fuel reprocessing plants

    SciTech Connect

    Burger, L.L.; Scheele, R.D.

    1983-07-01

    This report summarizes the status of radioiodine control in a nuclear fuel reprocessing plant with respect to capture, fixation, and disposal. Where possible, we refer the reader to a number of survey documents which have been published in the last four years. We provide updates where necessary. Also discussed are factors which must be considered in developing criteria for iodine control. For capture from gas streams, silver mordenite and a silver nitrate impregnated silica (AC-6120) are considered state-of-the-art and are recommended. Three aqueous scrubbing processes have been demonstrated: Caustic scrubbing is simple but probably will not give an adequate iodine retention by itself. Mercurex (mercuric nitrate-nitric acid scrubbing) has a number of disadvantages including the use of toxic mercury. Iodox (hyperazeotropic nitric acid scrubbing) is effective but employs a very corrosive and hazardous material. Other technologies have been tested but require extensive development. The waste forms recommended for long-term storage or disposal are silver iodide, the iodates of barium, strontium, or calcium, and silver loaded sorbents, all fixed in cement. Copper iodide in bitumen (asphalt) is a possibility but requires testing. The selection of a specific form will be influenced by the capture process used.

  13. Overview of reductants utilized in nuclear fuel reprocessing/recycling

    SciTech Connect

    Paviet-Hartmann, P.; Riddle, C.; Campbell, K.; Mausolf, E.

    2013-07-01

    The most widely used reductant to partition plutonium from uranium in the Purex process was ferrous sulfamate, other alternates were proposed such as hydrazine-stabilized ferrous nitrate or uranous nitrate, platinum catalyzed hydrogen, and hydrazine, hydroxylamine salts. New candidates to replace hydrazine or hydroxylamine nitrate (HAN) are pursued worldwide. They may improve the performance of the industrial Purex process towards different operations such as de-extraction of plutonium and reduction of the amount of hydrazine which will limit the formation of hydrazoic acid. When looking at future recycling technologies using hydroxamic ligands, neither acetohydroxamic acid (AHA) nor formohydroxamic acid (FHA) seem promising because they hydrolyze to give hydroxylamine and the parent carboxylic acid. Hydroxyethylhydrazine, HOC{sub 2}H{sub 4}N{sub 2}H{sub 3} (HEH) is a promising non-salt-forming reductant of Np and Pu ions because it is selective to neptunium and plutonium ions at room temperature and at relatively low acidity, it could serve as a replacement of HAN or AHA for the development of a novel used nuclear fuel recycling process.

  14. Laser ultrasonics evaluation and testing of coated HTR nuclear fuel

    NASA Astrophysics Data System (ADS)

    Amziane, Ahmed; Amari, Mohamed; Mounier, Denis; Breteau, Jean-Marc; Joly, Nicolas; Edely, Mathieu; Larcher, Maxime; Noiré, Paul; Banchet, Julien; Tisseur, David; Gusev, Vitalyi

    2011-05-01

    Laser ultrasonics was applied to the manufacturing control of the quality and integrity (no failure) of coated spherical particles designed for High Temperature Reactors (HTR). The coating of the nuclear fuel kernel is designed to prevent from the diffusion of fission products outside the particle during reactor operation. The quality assessment of the coating layers is of major importance. Using laser ultrasonics, we determined the vibration eigenmodes of dummy HTR particles. The vibration spectrum of a HTR particle provides a non-destructive method of evaluating some important mechanical parameters of the coating. Moreover, without damaging the particle, the laser ultrasonics technique allows to test the presence of a crack in the SiC layer, through the observation of the particle vibration spectrum, which is significantly changed, compared to that of a defect-free particle. We applied nanosecond acoustic pulses, i.e., high frequency laser-generated ultrasound, to measure the acoustic velocity of longitudinal waves the SiC layer. This technique provides an alternative method of evaluation of the Young modulus of the SiC layer. We measure the velocity of surface acoustic waves (SAW) on a pyrocarbon layer cross-section and we demonstrated that the anisotropy of the internal pyrocarbon layer can be evaluated by laser ultrasonics.

  15. Disposition of salt-waste from pyrochemical nuclear fuel processing

    SciTech Connect

    Vance, E.R.

    2007-07-01

    Waste salts from pyrochemical processing of nuclear fuel can be immobilised in sodalite if consolidated by hot isostatic pressing (HIP) at {approx}750 deg. C/100 MPa in thick stainless steel 316 cans. Other canning materials for this purpose also look possible. Spodiosite-based waste forms do not look promising in terms of leach resistance and their incorporation of alkali ions and compatibility with other phases which could potentially accommodate fission products, such as NaZr{sub 2}(PO{sub 4}){sub 3} or alumino-phosphate glass. Chloro- or fluor-apatite-based waste forms however have been reported to successfully accommodate fission products and alkalis which would be derived from either chloride- or fluoride-based waste pyro-processing salts. The presence of 10 or 20 wt% of additional Whitlockite, Ca{sub 3}(PO{sub 4}){sub 2}, should allow chemical flexibility to maintain the same qualitative phase assemblage when there are variations in the waste feed and in the waste/precursor ratios. Experimental verification of incorporation of the full complement of waste salts and fission products is not yet complete however. Apatite-rich samples could likely be HIPed in Inconel 600 cans. Other candidate HIP canning materials such as Alloy 22 or Inconel 625 are under study by encapsulating them in the candidate waste form and studying their interaction or otherwise with the waste form. (author)

  16. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    SciTech Connect

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J.; Brey, R.F.; Wright, R.N.; Windes, W.F.

    1999-09-03

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.

  17. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    SciTech Connect

    N. E. Pettit

    2001-07-13

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident.

  18. Status of the Canadian nuclear fuel waste management program

    SciTech Connect

    Allan, C.J.; Stephens, M.E. )

    1992-01-01

    The Canadian Concept for the permanent disposal of nuclear fuel waste has been developed extensively over the past several years, and is now well-advanced. The Concept, which involves the construction of a waste vault 500 to 1000 metres deep in plutonic rock located in the Canadian Precambrian Shield, is supported by an R D program with the following objectives: (1) to develop and demonstrate technology to site, design, build and operate a disposal facility; (2) to develop and demonstrate a methodology to evaluate the performance of the disposal system; and (3) to demonstrate that sites are likely to exist in the Canadian Precambrian Shield that would meet the regulatory requirements. A combination of engineered and natural barriers will be used to ensure that the vault design will meet rigorous safety standards. Experimental work is being carried out to elucidate all the important phenomena associated with the safety of the vault, including the performance of engineered barriers, natural geological barriers, and the biosphere.

  19. Deployment evaluation methodology for the electrometallurgical treatment of DOE-EM spent nuclear fuel

    SciTech Connect

    Dahl, C.A.; Adams, J.P.; Ramer, R.J.

    1998-07-01

    Part of the Department of Energy (DOE) spent nuclear fuel (SNF) inventory may require some type of treatment to meet acceptance criteria at various disposition sites. The current focus for much of this spent nuclear fuel is the electrometallurgical treatment process under development at Argonne National Laboratory. Potential flowsheets for this treatment process are presented. Deployment of the process for the treatment of the spent nuclear fuel requires evaluation to determine the spent nuclear fuel program need for treatment and compatibility of the spent nuclear fuel with the process. The evaluation of need includes considerations of cost, technical feasibility, process material disposition, and schedule to treat a proposed fuel. A siting evaluation methodology has been developed to account for these variables. A work breakdown structure is proposed to gather life-cycle cost information to allow evaluation of alternative siting strategies on a similar basis. The evaluation methodology, while created specifically for the electrometallurgical evaluation, has been written such that it could be applied to any potential treatment process that is a disposition option for spent nuclear fuel. Future work to complete the evaluation of the process for electrometallurgical treatment is discussed.

  20. Preliminary Evaluation of Removing Used Nuclear Fuel From Nine Shutdown Sites

    SciTech Connect

    Maheras, Steven J.; Best, Ralph; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul

    2013-04-30

    The Blue Ribbon Commission on America’s Nuclear Future 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. In this report, a preliminary evaluation of removing used nuclear fuel from nine shutdown sites was conducted. The shutdown sites included Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion. At these sites a total of 7649 used nuclear fuel assemblies and a total of 2813.2 metric tons heavy metal (MTHM) of used nuclear fuel are contained in 248 storage canisters. In addition, 11 canisters containing greater-than-Class C (GTCC) low-level radioactive waste are stored at these sites. The evaluation was divided in four components: • characterization of the used nuclear fuel and GTCC low-level radioactive waste inventory at the shutdown sites • an evaluation of the onsite transportation conditions at the shutdown sites • an evaluation of the near-site transportation infrastructure and experience relevant to the shipping of transportation casks containing used nuclear fuel from the shutdown sites • an evaluation of the actions necessary to prepare for and remove used nuclear fuel and GTCC low-level radioactive waste from the shutdown sites. Using these evaluations the authors developed time sequences of activities and time durations for removing the used nuclear fuel and GTCC low-level radioactive waste from a single shutdown site, from three shutdown sites located close to each other, and from all nine shutdown sites.

  1. A Practical Solution to Used Nuclear Fuel Treatment to Enable Sustained Nuclear Energy and Recovery of Vital Materials

    SciTech Connect

    Collins, Emory D; DelCul, Guillermo D; Rushton, James E; Williams, Kent Alan

    2010-01-01

    A systems analysis, based on previous in-depth studies, was made to develop an advanced used fuel recycling approach in which over 90% of the fuel components and cladding material are recovered and reused. This can minimize the waste requiring disposition to a geologic repository while enabling sustained nuclear energy and recovery of vital materials. The analysis assumed that: (1) continuing delays will occur in locating a site and licensing a geologic repository in the United States; (2) continued storage of used fuel at reactor sites or central storage locations is not a permanent solution; and (3) public perception is that the unresolved used fuel disposal problem is a deterrent to the continuation and expansion of the use of nuclear energy. The practical solution recommended is to move forward with a parallel approach to (1) deployment of industrialized recycling of used fuels which is a multi-decade process, and (2) focused R&D to recover the re-usable components and minimize residual waste. This approach would use concepts for proliferation resistant recycle facilities, processing oldest fuels first, and incorporating more-complete recycling of used fuel components. In addition to uranium and transuranic actinide recycle, the R&D would be focused on recovery and re-use of valuable components such as noble metals, lighter lanthanide elements, xenon gas, and zirconium from cladding. While the eventual need for a geologic repository will remain, the methods recommended can delay the need, minimize the capacity required, and significantly reduce the hazard of the wastes disposed. With no decision, the path forward for nuclear waste (used fuels) disposal remains uncertain, with many diverse technologies being considered. However, a decision to take advantage of proliferation resistant facility design, processing aged fuels, and incorporation of the concepts of nearcomplete recycling will provide the focus and path forward to a practical solution to the problem of

  2. Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel

    NASA Astrophysics Data System (ADS)

    Lopes, Denise A.; Zimmermann, Angelo J. O.; Silva, Selma L.; Piqueira, J. R. C.

    2016-05-01

    Uranium alloys in a monolithic form have been considered attractive candidates for high density nuclear fuel. However, this high-density fissile material configuration keeps the volume permitted for the retention of fission products at a minimum. Additionally, the monolithic nuclear fuel has a peculiar configuration, whereby the fuel is in direct contact with the cladding. How this fuel configuration will retain fission products and how this will affect its integrity under various physical conditions - such as thermal cycling - are some of the technological problems for this new fuel. In this paper, the effect of out-of-pile thermal cycling is studied for a monolithic fuel plate produced by a hot co-rolling method using U-10Mo (wt %) as the fuel alloy and Zircaloy-4 as the cladding material. After performing 10 thermal cycles from 25 to 400 °C at a rate of 1 °C/min (∼125 h), the fuel alloy presented several fractures that were observed to occur in the last three cycles. These cracks nucleated approximately in the center of the fuel alloy and crossed the interdiffusion zone initiating an internal crack in the cladding. The results suggest that the origin of these fractures is the thermal fatigue of the U-10Mo alloy caused due to the combination of two factors: (i) the high difference in the thermal expansion coefficient of the fuel and of the cladding material, and (ii) the bound condition of fuel/cladding materials in this fuel element configuration.

  3. Biokinetics of nuclear fuel compounds and biological effects of nonuniform radiation.

    PubMed Central

    Lang, S; Servomaa, K; Kosma, V M; Rytömaa, T

    1995-01-01

    Environmental releases of insoluble nuclear fuel compounds may occur at nuclear power plants during normal operation, after nuclear power plant accidents, and as a consequence of nuclear weapons testing. For example, the Chernobyl fallout contained extensive amounts of pulverized nuclear fuel composed of uranium and its nonvolatile fission products. The effects of these highly radioactive particles, also called hot particles, on humans are not well known due to lack of reliable data on the extent of the exposure. However, the biokinetics and biological effects of nuclear fuel compounds have been investigated in a number of experimental studies using various cellular systems and laboratory animals. In this article, we review the biokinetic properties and effects of insoluble nuclear fuel compounds, with special reference to UO2, PuO2, and nonvolatile, long-lived beta-emitters Zr, Nb, Ru, and Ce. First, the data on hot particles, including sources, dosimetry, and human exposure are discussed. Second, the biokinetics of insoluble nuclear fuel compounds in the gastrointestinal tract and respiratory tract are reviewed. Finally, short- and long-term biological effects of nonuniform alpha- and beta-irradiation on the gastrointestinal tract, lungs, and skin are discussed. Images p920-a Figure 1. PMID:8529589

  4. Significance of the Nuclear Fuel Cycle in the 21st Century

    SciTech Connect

    Nash, Kenneth L.; Lumetta, Gregg J.; Clark, Sue B.; Friese, Judah I.

    2006-07-31

    The combined effects of increasing industrialization around the world, the threat of global climate change, and decreasing availability of 'clean' fossil fuels will make the development of alternative energy sources more important in the coming decades. For fission-based nuclear power to contribute significantly to future energy supplies, it will be essential to maintain the improvements that have been made in plant operational efficiency, to license geological repositories for waste disposal, and to consider again the issue of recycling of spent nuclear fuels to recover its fuel value and to reduce the long-term radiotoxicity of the wastes. In this chapter, we present an overview of the nuclear fuel cycle from spent fuel recycling through the repository performance in the context of its importance to energy production in the 21st Century.

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

    SciTech Connect

    Koji Shirai

    2006-04-01

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

  6. Hazard & Operability Study for Removal of Spent Nuclear Fuel from the 324 Building

    SciTech Connect

    VAN KEUREN, J.C.

    2002-05-07

    A hazard and operability (HAZOP) study was conducted to examine the hazards associated with the removal of the spent nuclear fuel from the 324 Building. Fifty-nine potentially hazardous conditions were identified.

  7. Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities

    SciTech Connect

    MITCHELL, R.M.

    2000-10-12

    This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey.

  8. Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities

    SciTech Connect

    MITCHELL, R.M.

    2000-09-28

    This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey.

  9. Lead Slowing Down Spectrometry Analysis of Data from Measurements on Nuclear Fuel

    SciTech Connect

    Warren, Glen A.; Anderson, Kevin K.; Kulisek, Jonathan A.; Danon, Yaron; Weltz, Adam; Gavron, Victor A.; Harris, Jason; Stewart, Trevor N.

    2015-01-12

    Improved non-destructive assay of isotopic masses in used nuclear fuel would be valuable for nuclear safeguards operations associated with the transport, storage and reprocessing of used nuclear fuel. Our collaboration is examining the feasibility of using lead slowing down spectrometry techniques to assay the isotopic fissile masses in used nuclear fuel assemblies. We present the application of our analysis algorithms on measurements conducted with a lead spectrometer. The measurements involved a single fresh fuel pin and discrete 239Pu and 235U samples. We are able to describe the isotopic fissile masses with root mean square errors over seven different configurations to 6.35% for 239Pu and 2.7% for 235U over seven different configurations. Funding Source(s):

  10. Effective thermal conductivity method for predicting spent nuclear fuel cladding temperatures in a dry fill gas

    SciTech Connect

    Bahney, Robert

    1997-12-19

    This paper summarizes the development of a reliable methodology for the prediction of peak spent nuclear fuel cladding temperature within the waste disposal package. The effective thermal conductivity method replaces other older methodologies.

  11. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  12. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

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

    SciTech Connect

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    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.

  14. Method for forming nuclear fuel containers of a composite construction and the product thereof

    DOEpatents

    Cheng, Bo-Ching; Rosenbaum, Herman S.; Armijo, Joseph S.

    1984-01-01

    An improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof.

  15. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    SciTech Connect

    Edstrom, C.M.; Phillips, A.G.; Johnson, L.D.; Corle, R.R.

    1980-10-11

    The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels.

  16. A Historical Review of the Safe Transport of Spent Nuclear Fuel, Rev. 1

    SciTech Connect

    Connolly, Kevin J.; Pope, Ronald

    2016-09-01

    This report is a revision to M3 milestone M3FT-16OR090402028 for the former Nuclear Fuels Storage and Transportation Planning Project (NFST), “Safety Record of SNF Shipments.” The US Department of Energy (DOE) has since established the Office of Integrated Waste Management (IWM), which builds on the work begun by NFST, to develop an integrated waste management system for spent nuclear fuel (SNF), including the developm

  17. Potential opportunities for nano materials to help enable enhanced nuclear fuel performance

    SciTech Connect

    McClellan, Kenneth J.

    2012-06-06

    This presentation is an overview of the technical challenges for development of nuclear fuels with enhanced performance and accident tolerance. Key specific aspects of improved fuel performance are noted. Examples of existing nanonuclear projects and concepts are presented and areas of potential focus are suggested. The audience for this presentation includes representatives from: DOE-NE, other national laboratories, industry and academia. This audience is a mixture of nanotechnology experts and nuclear energy researchers and managers.

  18. Nuclear power generation and fuel cycle report 1997

    SciTech Connect

    1997-09-01

    Nuclear power is an important source of electric energy and the amount of nuclear-generated electricity continued to grow as the performance of nuclear power plants improved. In 1996, nuclear power plants supplied 23 percent of the electricity production for countries with nuclear units, and 17 percent of the total electricity generated worldwide. However, the likelihood of nuclear power assuming a much larger role or even retaining its current share of electricity generation production is uncertain. The industry faces a complex set of issues including economic competitiveness, social acceptance, and the handling of nuclear waste, all of which contribute to the uncertain future of nuclear power. Nevertheless, for some countries the installed nuclear generating capacity is projected to continue to grow. Insufficient indigenous energy resources and concerns over energy independence make nuclear electric generation a viable option, especially for the countries of the Far East.

  19. THE ATTRACTIVENESS OF MATERIAS ASSOCIATED WITH THORIUM-BASED NUCLEAR FUEL CYCLES FOR PHWRS

    SciTech Connect

    Prichard, Andrew W.; Niehus, Mark T.; Collins, Brian A.; Bathke, Charles G.; Ebbinghaus, Bartley B.; Hase, Kevin R.; Sleaford, Brad W.; Robel, Martin; Smith, Brian W.

    2011-07-17

    This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with thorium based nuclear fuel cycles. Specifically, this paper examines a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of natural uranium/233U/thorium. This paper uses a PHWR fueled with natural uranium as a base fuel cycle, and then compares material attractiveness of fuel cycles that use 233U/thorium salted with natural uranium. The results include the material attractiveness of fuel at beginning of life (BoL), end of life (EoL), and the number of fuel assemblies required to collect a bare critical mass of plutonium or uranium. This study indicates what is required to render the uranium as having low utility for use in nuclear weapons; in addition, this study estimates the increased number of assemblies required to accumulate a bare critical mass of plutonium that has a higher utility for use in nuclear weapons. This approach identifies that some fuel cycles may be easier to implement the International Atomic Energy Agency (IAEA) safeguards approach and have a more effective safeguards by design outcome. For this study, approximately one year of fuel is required to be reprocessed to obtain one bare critical mass of plutonium. Nevertheless, the result of this paper suggests that all spent fuel needs to be rigorously safeguarded and provided with high levels of physical protection. This study was performed at the request of the United States Department of Energy /National Nuclear Security Administration (DOE/NNSA). The methodology and key findings will be presented.

  20. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    SciTech Connect

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.