Science.gov

Sample records for nuclear fuel behaviour

  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. 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.

  3. Nuclear fuel cycle costs

    SciTech Connect

    Burch, W.D.; Haire, M.J.; Rainey, R.H.

    1982-02-01

    The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel cycle costs are given for the pressurized water reactor once-through and fuel recycle systems, and for the liquid-metal fast breeder reactor system. These calculations show that fuel cycle costs are a small part of the total power costs. For breeder reactors, fuel cycle costs are about half that of the present once-through system. The total power cost of the breeder reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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)

  9. 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.

  10. 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

  11. 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.

  12. 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%.

  13. 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.

  14. 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.

  15. 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.

  16. 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)

  17. Nuclear fuel electrorefiner

    SciTech Connect

    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.

  18. Swelling-resistant nuclear fuel

    DOEpatents

    Arsenlis, Athanasios; Satcher, Jr., Joe; Kucheyev, Sergei O.

    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.

  19. Inert matrix fuel behaviour in test irradiations

    NASA Astrophysics Data System (ADS)

    Hellwig, Ch.; Streit, M.; Blair, P.; Tverberg, T.; Klaassen, F. C.; Schram, R. P. C.; Vettraino, F.; Yamashita, T.

    2006-06-01

    Among others, three large irradiation tests on inert matrix fuels have been performed during the last five years: the two irradiation tests IFA-651 and IFA-652 in the OECD Halden Material Test Reactor and the OTTO irradiation in the High Flux Reactor in Petten. While the OTTO irradiation is already completed, the other two irradiations are still ongoing. The objectives of the experiments differ: for OTTO, the focus was on the comparison of different concepts of IMF, i.e. homogeneous fuel versus different types of heterogeneous fuel. In IFA-651, single phase yttria stabilized zirconia (YSZ) doped with Pu is compared with MOX. In IFA-652, the potential of calcia stabilized zirconia (CSZ) as a matrix with and without thoria is evaluated. The design of the three experiments is explained and the current status is reviewed. The experiments show that the homogeneous, single phase YSZ-based or CSZ-based fuel show good and stable irradiation behaviour. It can be said that homogeneous stabilized zirconia based fuel is the most promising IMF concept for an LWR environment. Nevertheless, the fuel temperatures were relatively high due to the low thermal conductivity, potentially leading to high fission gas release, and must be taken into account in the fuel design.

  20. Uranium thorium hydride nuclear fuel

    SciTech Connect

    Simnad, M.T.

    1985-01-15

    A nuclear fuel includes uranium dispersed within a thorium hydride matrix. The uranium may be in the form of particles including fissile and non-fissile isotopes. Various hydrogen to thorium ratios may be included in the matrix. The matrix with the fissile dispersion may be used as a complete fuel for a metal hydride reactor or may be combined with other fuels.

  1. 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.

  2. 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.

  3. Signature surveillance of nuclear fuel

    SciTech Connect

    Bernatowicz, H.; Schoenig, F.C.

    1982-08-31

    Typical nuclear fuel material contains tramp ferromagnetic particles of random size and distribution. Also, selected amounts of paramagnetic or ferromagnetic material can be added at random or at known positions in the fuel material. The fuel material in its nonmagnetic container can be scanned by magnetic susceptibility change detecting apparatus to provide a unique signal waveform of the container of fuel material as a signature thereof. At subsequent times in its life, the container similarly can be scanned to provide subsequent signatures. Comparison of the signatures reveals any alteration or tampering with the fuel material.

  4. Nuclear fuels status

    NASA Technical Reports Server (NTRS)

    Kania, Michael

    1991-01-01

    A discussion on coated particle fuel performance from a modular High Temperature Gas Reactor (HTGR) is presented along with experimental results. The following topics are covered: (1) the coated particle fuel concept; (2) the functional requirements; (3) performance limiting mechanisms; (4) fuel performance; and (5) methods/techniques for characterizing performance.

  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. 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.

  7. 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

  8. Nuclear reactors and the nuclear fuel cycle

    SciTech Connect

    Pearlman, H

    1989-11-01

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

  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 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.

  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 ELEMENTS FOR NUCLEAR REACTORS

    DOEpatents

    Blainey, A.; Lloyd, H.

    1961-07-11

    A method of sheathing a tubular fuel element for a nuclear reactor is described. A low melting metal core member is centered in a die, a layer of a powdered sheathing substance is placed on the bottom of the die, the tubular fuel element is inserted in the die, the space between the tubular fuel element and the die walls and core member is filled with the same powdered sheathing substance, a layer of the same substance is placed over the fissile material, and the charge within the die is subjected to pressure in the direction of the axis of the fuel element at the sintering temperature of the protective substance.

  13. 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.

  14. 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.

  15. 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.

  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. 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.

  18. 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)

  19. 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)

  20. 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.

  1. 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.

  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. Nuclear propulsion technology advanced fuels technology

    NASA Technical Reports Server (NTRS)

    Stark, Walter A., Jr.

    1993-01-01

    Viewgraphs on advanced fuels technology are presented. Topics covered include: nuclear thermal propulsion reactor and fuel requirements; propulsion efficiency and temperature; uranium fuel compounds; melting point experiments; fabrication techniques; and sintered microspheres.

  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. 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.

  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

    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.

  7. 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.

  8. Spent nuclear fuel sampling strategy

    SciTech Connect

    Bergmann, D.W.

    1995-02-08

    This report proposes a strategy for sampling the spent nuclear fuel (SNF) stored in the 105-K Basins (105-K East and 105-K West). This strategy will support decisions concerning the path forward SNF disposition efforts in the following areas: (1) SNF isolation activities such as repackaging/overpacking to a newly constructed staging facility; (2) conditioning processes for fuel stabilization; and (3) interim storage options. This strategy was developed without following the Data Quality Objective (DQO) methodology. It is, however, intended to augment the SNF project DQOS. The SNF sampling is derived by evaluating the current storage condition of the SNF and the factors that effected SNF corrosion/degradation.

  9. Closing the nuclear fuel cycle

    SciTech Connect

    Wilcox, P.

    1993-12-31

    Generally the case for closing the nuclear fuel cycle is based on the strategic value of the uranium and plutonium recovered by reprocessing spent fuel. The energy content of 1 t of spent fuel varies from 10,000 to 40,000 t of coal equivalent depending on the reactor type from which the spent fuel arises. Recycling in fast reactors would increase these values by a factor or roughly 40. Reprocessing in the UK has its roots in the technology developed during and after the 1939-45 war to provide plutonium for defence purposes. At BNFL`s Sellafield site in northern England the commercial reprocessing of spent fuel has been undertaken for over 30 years with a cumulative throughput of over 30,000 tU. Over 15,000 tU of the uranium recovered has been recycled and some 70% of all the UK`s AGR fuel has been produced from this material. As a consequence the UK`s bill for imported uranium has been reduced by several hundred million pounds sterling. This report discusses issues associated with reprocessing, uranium, and plutonium recycle.

  10. 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)

  11. Nuclear fuel, refueling, fuel handling, and licensing and regulation. Volume eleven

    SciTech Connect

    Not Available

    1986-01-01

    Volume eleven covers nuclear fuel (what is nuclear fuel, the nuclear fuel cycle, uranium mining, milling, and refining, uranium enrichment, nuclear fuel fabrication, fuel reprocessing), refueling and fuel handling (fuel assembly identification, fuel handling equipment, the fueling and refueling process, PWR refueling, BWR refueling), and licensing and regulation requirements (development of nuclear energy, federal licensing and regulatory organization, schedule for nuclear power plants, contents of reports to the Federal regulatory agency, nuclear power plant operator qualification).

  12. Corrosion of Nuclear Fuel Inside a Failed Copper Nuclear Waste Container

    SciTech Connect

    Broczkowski, Michael E.; Goldik, Jonathan S.; Santos, Billy G.; Noel, James J.; Shoesmith, David

    2007-07-01

    Canada's Nuclear Waste Management Organization has recommended to the Canadian federal government an adaptive phased management approach to the long-term management of used nuclear fuel. This approach includes isolation in a deep geologic repository. In such a repository, the fuel would be sealed inside a carbon steel-lined copper container. To assist the development of performance assessment models studies of fuel behaviour inside a failed waste container are underway. Using an iterative modeling and experimental approach, the important features and processes that determine fuel behaviour have been identified and studied. These features and processes are discussed and the results of studies to elucidate specific mechanisms and determine important parameter values summarized. (authors)

  13. 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.

  14. A mechanistic code for intact and defective nuclear fuel element performance

    NASA Astrophysics Data System (ADS)

    Shaheen, Khaled

    During reactor operation, nuclear fuel elements experience an environment featuring high radiation, temperature, and pressure. Predicting in-reactor performance of nuclear fuel elements constitutes a complex multi-physics problem, one that requires numerical codes to be solved. Fuel element performance codes have been developed for different reactor and fuel designs. Most of these codes simulate fuel elements using one-or quasi-two-dimensional geometries, and some codes are only applicable to steady state but not transient behaviour and vice versa. Moreover, while many conceptual and empirical separate-effects models exist for defective fuel behaviour, wherein the sheath is breached allowing coolant ingress and fission gas escape, there have been few attempts to predict defective fuel behaviour in the context of a mechanistic fuel performance code. Therefore, a mechanistic fuel performance code, called FORCE (Fuel Operational peRformance Computations in an Element) is proposed for the time-dependent behaviour of intact and defective CANDU nuclear fuel elements. The code, which is implemented in the COMSOL Multiphysics commercial software package, simulates the fuel, sheath, and fuel-to-sheath gap in a radial-axial geometry. For intact fuel performance, the code couples models for heat transport, fission gas production and diffusion, and structural deformation of the fuel and sheath. The code is extended to defective fuel performance by integrating an adapted version of a previously developed fuel oxidation model, and a model for the release of radioactive fission product gases from the fuel to the coolant. The FORCE code has been verified against the ELESTRES-IST and ELESIM industrial code for its predictions of intact fuel performance. For defective fuel behaviour, the code has been validated against coulometric titration data for oxygen-to-metal ratio in defective fuel elements from commercial reactors, while also being compared to a conceptual oxidation model

  15. Variants of closing the nuclear fuel cycle

    NASA Astrophysics Data System (ADS)

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

    2015-12-01

    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.

  16. 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.

  17. 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.

  18. Spent nuclear fuel disposal liability insurance

    SciTech Connect

    Martin, D.W.

    1984-01-01

    This thesis examines the social efficiency of nuclear power when the risks of accidental releases of spent fuel radionuclides from a spent fuel disposal facility are considered. The analysis consists of two major parts. First, a theoretical economic model of the use of nuclear power including the risks associated with releases of radionuclides from a disposal facility is developed. Second, the costs of nuclear power, including the risks associated with a radionuclide release, are empirically compared to the costs of fossil fuel-fired generation of electricity. Under the provisions of the Nuclear Waste Policy Act of 1982, the federally owned and operated spent nuclear fuel disposal facility is not required to maintain a reserve fund to cover damages from an accidental radionuclide release. Thus, the risks of a harmful radionuclide release are not included in the spent nuclear fuel disposal fee charged to the electric utilities. Since the electric utilities do not pay the full, social costs of spent fuel disposal, they use nuclear fuel in excess of the social optimum. An insurance mechanism is proposed to internalize the risks associated with spent fueled disposal. Under this proposal, the Federal government is required to insure the disposal facility against any liabilities arising from accidental releases of spent fuel radionuclides.

  19. Spent Nuclear Fuel (SNF) Project Execution Plan

    SciTech Connect

    LEROY, P.G.

    2000-11-03

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

  20. 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.

  1. Modeling and Simulation of Nuclear Fuel Materials

    SciTech Connect

    Devanathan, Ram; Van Brutzel, Laurent; Tikare, Veena; Bartel, Timothy; Besmann, Theodore M; Stan, Marius; Van Uffelen, Paul

    2010-01-01

    We review the state of modeling and simulation of nuclear fuels with emphasis on the most widely used nuclear fuel, UO2. The hierarchical scheme presented represents a science-based approach to modeling nuclear fuels by progressively passing information in several stages from ab initio to continuum levels. Such an approach is essential to overcome the challenges posed by radioactive materials handling, experimental limitations in modeling extreme conditions and accident scenarios and small time and distance scales of fundamental defect processes. When used in conjunction with experimental validation, this multiscale modeling scheme can provide valuable guidance to development of fuel for advanced reactors to meet rising global energy demand.

  2. Modeling and Simulation of Nuclear Fuel Materials

    SciTech Connect

    Devanathan, Ramaswami; Van Brutzel, Laurent; Chartier, Alan; Gueneau, Christine; Mattsson, Ann E.; Tikare, Veena; Bartel, Timothy; Besmann, T. M.; Stan, Marius; Van Uffelen, Paul

    2010-10-01

    We review the state of modeling and simulation of nuclear fuels with emphasis on the most widely used nuclear fuel, UO2. The hierarchical scheme presented represents a science-based approach to modeling nuclear fuels by progressively passing information in several stages from ab initio to continuum levels. Such an approach is essential to overcome the challenges posed by radioactive materials handling, experimental limitations in modeling extreme conditions and accident scenarios, and the small time and distance scales of fundamental defect processes. When used in conjunction with experimental validation, this multiscale modeling scheme can provide valuable guidance to development of fuel for advanced reactors to meet rising global energy demand.

  3. Nuclear fuels for very high temperature applications

    SciTech Connect

    Lundberg, L.B.; Hobbins, R.R.

    1992-08-01

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO{sub 2} or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

  4. Nuclear fuels for very high temperature applications

    SciTech Connect

    Lundberg, L.B.; Hobbins, R.R.

    1992-01-01

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO{sub 2} or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

  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. Simulated nuclear reactor fuel assembly

    DOEpatents

    Berta, V.T.

    1993-04-06

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

  7. Simulated nuclear reactor fuel assembly

    DOEpatents

    Berta, Victor T.

    1993-01-01

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

  8. 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.

  9. 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.

  10. 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)

  11. 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.

  12. 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)

  13. 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.

  14. Spent Nuclear Fuel Project dose management plan

    SciTech Connect

    Bergsman, K.H.

    1996-03-01

    This dose management plan facilitates meeting the dose management and ALARA requirements applicable to the design activities of the Spent Nuclear Fuel Project, and establishes consistency of information used by multiple subprojects in ALARA evaluations. The method for meeting the ALARA requirements applicable to facility designs involves two components. The first is each Spent Nuclear Fuel Project subproject incorporating ALARA principles, ALARA design optimizations, and ALARA design reviews throughout the design of facilities and equipment. The second component is the Spent Nuclear Fuel Project management providing overall dose management guidance to the subprojects and oversight of the subproject dose management efforts.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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)

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Pressurized heavy water reactor fuel behaviour in power ramp conditions

    NASA Astrophysics Data System (ADS)

    Ionescu, S.; Uţă, O.; Pârvan, M.; Ohâi, D.

    2009-03-01

    In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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...

  11. Immobilization and geological disposal of nuclear fuel waste.

    PubMed

    Tait, J C

    1984-08-01

    The Canadian Nuclear Fuel Waste Management Program is developing methods for the safe disposal of both used nuclear fuel and fuel recycle waste. The disposal strategy is based on interim storage of the used fuel, immobilization of either used fuel or recycle waste, and disposal, deep in a stable geological formation in the Canadian Shield. The disposal concept proposes a multibarrier system to inhibit the release of the radioactive waste from the disposal vault. The principal components of the multibarrier system are (i) the waste form in which the radionuclides are immobilized, (ii) engineered barriers including high integrity containers, buffers and backfills designed to retard the movement of groundwaters in the disposal vault, and (iii) the natural barrier provided by the massive geological formation itself. The research programs to investigate this concept are discussed briefly. Several different waste forms are being developed for the immobilization of high-level fuel recycle waste, including glass, glass-ceramics and crystalline materials. Dissolution of these materials in groundwater is the only likely scenario that could lead to radionuclide release. The factors that influence the aqueous dissolution behaviour of these materials are reviewed. PMID:6488089

  12. 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)

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Plumb nozzle for nuclear fuel assembly

    SciTech Connect

    Silverblatt, B.L.

    1986-02-25

    An elongated nuclear reactor fuel assembly is described having an asymmetric weight distribution across a cross section, including a nozzle affixed to one end of the fuel assembly having lifting surfaces formed thereon on which the fuel assembly can be supported when suspended from the surfaces. At least one of the lifting surfaces is located at a first elevation relative to the longitudinal axis of the fuel assembly. A second of the lifting surfaces is located at a second elevation different from the first elevation wherein the difference in the first and second elevations is sized to offset the asymmetric weight distribution when the fuel assembly is supported from the first and second surface so that when so supported the fuel assembly will hang plumb.

  1. 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.

  2. 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…

  3. Insights on fission products behaviour in nuclear severe accident conditions by X-ray absorption spectroscopy

    NASA Astrophysics Data System (ADS)

    Geiger, E.; Bès, R.; Martin, Ph; Pontillon, Y.; Ducros, G.; Solari, P. L.

    2016-04-01

    Many research programs have been carried out aiming to understand the fission products behaviour during a Nuclear Severe Accident. Most of these programs used highly radioactive irradiated nuclear fuel, which requires complex instrumentation. Moreover, the radioactive character of samples hinders an accurate chemical characterisation. In order to overcome these difficulties, SIMFUEL stand out as an alternative to perform complementary tests. A sample made of UO2 doped with 11 fission products was submitted to an annealing test up to 1973 K in reducing atmosphere. The sample was characterized before and after the annealing test using SEM-EDS and XAS at the MARS beam-line, SOLEIL Synchrotron. It was found that the overall behaviour of several fission products (such as Mo, Ba, Pd and Ru) was similar to that observed experimentally in irradiated fuels and consistent with thermodynamic estimations. The experimental approach presented in this work has allowed obtaining information on chemical phases evolution under nuclear severe accident conditions, that are yet difficult to obtain using irradiated nuclear fuel samples.

  4. Insights on fission products behaviour in nuclear severe accident conditions by X-ray absorption spectroscopy

    NASA Astrophysics Data System (ADS)

    Geiger, E.; Bès, R.; Martin, Ph; Pontillon, Y.; Ducros, G.; Solari, P. L.

    2016-04-01

    Many research programs have been carried out aiming to understand the fission products behaviour during a Nuclear Severe Accident. Most of these programs used highly radioactive irradiated nuclear fuel, which requires complex instrumentation. Moreover, the radioactive character of samples hinders an accurate chemical characterisation. In order to overcome these difficulties, SIMFUEL stand out as an alternative to perform complementary tests. A sample made of UO2 doped with 11 fission products was submitted to an annealing test up to 1973 K in reducing atmosphere. The sample was characterized before and after the annealing test using SEM-EDS and XAS at the MARS beam-line, SOLEIL Synchrotron. It was found that the overall behaviour of several fission products (such as Mo, Ba, Pd and Ru) was similar to that observed experimentally in irradiated fuels and consistent with thermodynamic estimations. The experimental approach presented in this work has allowed obtaining information on chemical phases evolution under nuclear severe accident conditions, that are yet difficult to obtain using irradiated nuclear fuel samples.

  5. Intelligent Automated Nuclear Fuel Pellet Inspection System

    SciTech Connect

    S. Keyvan

    1999-11-01

    At the present time, nuclear pellet inspection is performed manually using naked eyes for judgment and decisionmaking on accepting or rejecting pellets. This current practice of pellet inspection is tedious and subject to inconsistencies and error. Furthermore, unnecessary re-fabrication of pellets is costly and the presence of low quality pellets in a fuel assembly is unacceptable. To improve the quality control in nuclear fuel fabrication plants, an automated pellet inspection system based on advanced techniques is needed. Such a system addresses the following concerns of the current manual inspection method: (1) the reliability of inspection due to typical human errors, (2) radiation exposure to the workers, and (3) speed of inspection and its economical impact. The goal of this research is to develop an automated nuclear fuel pellet inspection system which is based on pellet video (photographic) images and uses artificial intelligence techniques.

  6. 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.

  7. 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.

  8. 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.

  9. Safeguarding and Protecting the Nuclear Fuel Cycle

    SciTech Connect

    Trond Bjornard; Humberto Garcia; William Desmond; Scott Demuth

    2010-11-01

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

  10. 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.

  11. 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.

  12. Nuclear fuel elements made from nanophase materials

    SciTech Connect

    Heubeck, Norman B.

    1997-12-01

    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 3,000 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.

  13. 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.

  14. 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.

  15. 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.

  16. World nuclear fuel market. Seventeenth annual meeting

    SciTech Connect

    Not Available

    1990-01-01

    The papers presented at the seventeenth World Nuclear Fuels Market meeting are cataloged individually. This volume includes information on the following areas of interest: historical and current aspects of the uranium and plutonium market with respect to supply and demand, pricing, spot market purchasing, and other market phenomena; impact of reprocessing and recycling uranium, plutonium, and mixed oxide fuels; role of individual countries in the market: Hungary, Germany, the Soviet Union, Czechoslovakia, France, and the US; the impact of public opinion and radioactive waste management on the nuclear industry, and a debate regarding long term versus short term contracting by electric utilities for uranium and enrichment services.

  17. 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)

  18. Locking support for nuclear fuel assemblies

    DOEpatents

    Ledin, Eric

    1980-01-01

    A locking device for supporting and locking a nuclear fuel assembly within a cylindrical bore formed by a support plate, the locking device including a support and locking sleeve having upwardly extending fingers forming wedge shaped contact portions arranged for interaction between an annular tapered surface on the fuel assembly and the support plate bore as well as downwardly extending fingers having wedge shaped contact portions arranged for interaction between an annularly tapered surface on the support plate bore and the fuel assembly whereby the sleeve tends to support and lock the fuel assembly in place within the bore by its own weight while facilitating removal and/or replacement of the fuel assembly.

  19. Nuclear reactor spent fuel storage rack

    SciTech Connect

    Machado, O.J.; Flynn, W.M.; Flanders, H.E. Jr.; Booker, L.W.

    1989-04-11

    A fuel rack is described for use in storing nuclear fuel assemblies in a nuclear fuel storage pool having a floor on which an upwardly projecting stud is mounted; the fuel rack comprising: a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, the base-plate means supporting a first network of interlaced beams which form a multiplicity of polygonal openings; a second network of interlaced beams forming polygonal openings positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, each cell positioned in a corresponding pair of the aligned polygonal openings, each cell being open at both ends with a guiding funnel at the upper end, and the cells being positioned over the flow openings in the base-plate to permit flow of coolant through the cells; spaced, outwardly directed, projections on the outer surfaces of the sides of the cells near the tops and bottoms of the sides thereof, each cell being sized to be received within a corresponding of the pair of aligned polygonal openings in which the cells are respectively positioned; and means fixedly securing the projections to the beams in the first and second networks of beams thereby to provide a substantially rigid fuel rack of modular design.

  20. 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.

  1. World nuclear fuel market. Eighteenth annual meeting

    SciTech Connect

    Not Available

    1991-01-01

    The papers presented at the eighteenth World Nuclear Fuels Market meeting are cataloged separately. This volume includes information on the following areas of interest: world uranium enrichment capacity and enriched uranium inventories; the impact of new enrichment technologies; predictions of future market trends; non-proliferation aspects of nuclear trade; and a debate as to whether uranium can be successfully traded on a commodities exchange.

  2. 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.

  3. Nuclear Propulsion using Thin Foiled Fuel

    NASA Astrophysics Data System (ADS)

    Takahashi, H.

    1998-11-01

    A new way to produce plasma for nuclear propulsion is proposed. A thin foiled fuel can be used for converting fission energy to propulsion energy efficiently. The fission products coming out of the thin foil directly ionize the hydrogen molecules which are used for propulsion. Thus very small portion of fission energy deposited in the thin foil, and integrity of the thin foiled fuel can be maintained even in high nuclear power. Fuel material with large thermal fission cross-section is preferable to make thin foiled fuel and the heat deposition in the foil can be reduced. To get high power from the foiled fuel assembly, thermal neutrons which are created out from the assembly can be supplied, or the assembly itself can create the high intensity thermal neutrons by self-multiplication. A flexible design of a highly efficient nuclear propulsion system can be made. The thickness of the foil and the maintenance of the thermo-mechanical integrity can be determined from the fission cross-section and the slowing down power for fission products. The talk discusses the issues related to heat removal from the assembly.

  4. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    SciTech Connect

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

    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.

  5. Spent Nuclear Fuel (SNF) Project Product Specification

    SciTech Connect

    PAJUNEN, A.L.

    2000-01-20

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project.

  6. 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)

  7. Spent nuclear fuel project product specification

    SciTech Connect

    PAJUNEN, A.L.

    1999-02-25

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Spent nuclear fuel project product specification

    SciTech Connect

    Pajunen, A.L.

    1998-01-30

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification.

  13. 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.

  14. Thermodynamic calculations of oxygen self-diffusion in mixed-oxide nuclear fuels

    DOE PAGESBeta

    Parfitt, David C.; Cooper, Michael William; Rushton, Michael J.D.; Christopoulos, S. R.; Fitzpatrick, M. E.; Chroneos, A.

    2016-07-29

    Mixed-oxide fuels containing uranium with thorium and/or plutonium may play an important part in future nuclear fuel cycles. There are, however, significantly less data available for these materials than conventional uranium dioxide fuel. In the present study, we employ molecular dynamics calculations to simulate the elastic properties and thermal expansivity of a range of mixed oxide compositions. These are then used to support equations of state and oxygen self-diffusion models to provide a self-consistent prediction of the behaviour of these mixed oxide fuels at arbitrary compositions.

  15. VISION - Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics

    SciTech Connect

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

    2006-02-01

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

  16. 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.

  17. Nuclear fuel assembly with coolant conducting tube

    SciTech Connect

    Dunlap, T. G.; Cearley, J. E.; Jameson, W. G. Jr.; Mefford, C. R.; Nelson, H. L.

    1983-12-13

    In a nuclear fuel assembly having a coolant conducting or water tube which also retains the spacers in axial position, the fuel rods experience greater axial growth with exposure than the water tube creating a risk that the water tube might become disengaged from the supporting tie plates. An arrangement for preventing such disengagement is described including lengthened end plug shanks for the water tube, a protective boss surrounding the lower end plug shank to protect it from flow induced vibration, a conical seat for the lower end plug and an arrangement for limiting upward movement of the water tube.

  18. Use of cermet fueled nuclear reactors for direct nuclear propulsion

    SciTech Connect

    Bhattacharyya, S.K.; Carlson, L.W.; Kuczen, K.D.; Hanan, N.A.; Palmer, R.G.; Von Hoomissen, J.; Chiu, W.; Haaland, R.

    1988-07-01

    There has been a renewal of interest in Direct Nuclear Propulsion (DNP) because of the Air Force Forecast II recommendation for the development of the technology. Several nuclear concepts have been proposed to meet the Direct Nuclear Propulsion challenge. In this paper we will present results of an initial study of the potential of a cermet fueled nuclear system in providing the desired DNP capabilities and featuring a set of unique safety characteristics. The concept of cermet fuel for DNP applications was first developed by ANL and GE working independently more than 20 years ago. The two organizations came to several remarkably consistent conclusions. The present work has consisted of collecting a unified set of design parameters from the set of design results produced in the earlier work. The conclusion of this exercise was that a cermet-fueled DNP design looked extremely promising from performance and safety considerations and that it deserves serious consideration when the decision to develop one or more concepts for DNP is made.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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)

  6. Current Comparison of Advanced Nuclear Fuel Cycles

    SciTech Connect

    Steven Piet; Trond Bjornard; Brent Dixon; Robert Hill; Gretchen Matthern; David Shropshire

    2007-04-01

    This paper compares potential nuclear fuel cycle strategies – once-through, recycling in thermal reactors, sustained recycle with a mix of thermal and fast reactors, and sustained recycle with fast reactors. Initiation of recycle starts the draw-down of weapons-usable material and starts accruing improvements for geologic repositories and energy sustainability. It reduces the motivation to search for potential second geologic repository sites. Recycle in thermal-spectru

  7. 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.

  8. Report on interim storage of spent nuclear fuel

    SciTech Connect

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  9. 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.

  10. 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.

  11. A model for predicting coolant activity behaviour for fuel-failure monitoring analysis

    NASA Astrophysics Data System (ADS)

    El-Jaby, Ali

    A CANDU fuel element becomes defective when the Zircaloy-4 sheath is breached, allowing high pressure heavy water (D2O) coolant to enter the fuel-to-sheath gap, thereby creating a direct path for fission products (mainly volatile species of iodine and noble gases) and fuel debris to escape into the primary heat transport system (PHTS). In addition, the entry of D 2O coolant into the fuel-to-sheath gap may cause the UO2 fuel to oxidize, which in turn can augment the rate of fission product release into the PHTS. The release of fission products and fuel debris into the PHTS will elevate circuit contamination levels, consequently increasing radiation exposure to station personnel during maintenance tasks. Moreover, the continued operation of a defective fuel element may diminish its thermal performance due to fuel oxidation effects. It is therefore desirable to discharge defective fuel as soon as possible. Hence, a better understanding of defective fuel behaviour is required in order to develop an improved methodology for fuel-failure monitoring and PHTS coolant activity prediction. A mathematical model has been developed to predict the release of volatile fission products from operating defective nuclear fuel elements. The fission product activity in both the fuel-to-sheath gap and PHTS coolant as a function of time can be predicted during all reactor operations including steady-state operation as well as reactor shutdown, startup, and bundle-shifting manoeuvres. In addition, an improved ability to predict the PHTS coolant activity of the 135Xe isotope in commercial reactors is discussed. Moreover, a method to approximate both the burnup and the amount of the tramp uranium deposits in-core, as well as the tramp uranium fission rate is proposed. The model has been implemented as the STAR (Steady-state and Transient Activity Release) stand-alone code written in the C++ programming language using a custom developed finite-difference variable-mesh (FDVM) numerical

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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

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

    NASA Astrophysics Data System (ADS)

    Settle, Frank A.

    2009-03-01

    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 dominates all other aspects of the nuclear fuel cycle. An understanding of this chemistry is necessary to address the economic, environmental, safety, and proliferation issues that are essential to any substantive evaluation of nuclear power's contribution to the global energy portfolio. This article describes the role of chemistry in each component of the cycle from the metallurgy of uranium to the disposition of spent reactor fuel. It also addresses the economics of the components of the cycle and the costs of nuclear power relative to other sources of energy.

  2. 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.

  3. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    SciTech Connect

    F. Habashi

    1998-06-26

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

  4. Fission products behaviour in UO2 submitted to nuclear severe accident conditions

    NASA Astrophysics Data System (ADS)

    Geiger, E.; Bès, R.; Martin, P.; Pontillon, Y.; Solari, P. L.; Salome, M.

    2016-05-01

    The objective of this work was to study the molybdenum chemistry in UO2 based materials, known as SIMFUELS. These materials could be used as an alternative to irradiated nuclear fuels in the study of fission products behaviour during a nuclear severe accident. UO2 samples doped with 12 stable isotopes of fission products were submitted to annealing tests in conditions representative to intermediate steps of severe accidents. Samples were characterized by SEM-EDS and XAS. It was found that Mo chemistry seems to be more complex than what is normally estimated by thermodynamic calculations: XAS spectra indicate the presence of Mo species such as metallic Mo, MoO2, MoO3 and Cs2MoO4.

  5. 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.

  6. 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.

  7. CHARACTERIZATION OF HYDROGEN CONTENT IN ZIRCALOY-4 NUCLEAR FUEL CLADDING

    SciTech Connect

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

    2010-02-22

    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.

  8. The behaviour under irradiation of molybdenum matrix for inert matrix fuel containing americium oxide (CerMet concept)

    NASA Astrophysics Data System (ADS)

    D'Agata, E.; Knol, S.; Fedorov, A. V.; Fernandez, A.; Somers, J.; Klaassen, F.

    2015-10-01

    Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors or Accelerator Driven System (ADS, subcritical reactors dedicated to transmutation) of long-lived nuclides like 241Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. In order to safely burn americium in a fast reactor or ADS, it must be incorporated in a matrix that could be metallic (CerMet target) or ceramic (CerCer target). One of the most promising matrix to incorporate Am is molybdenum. In order to address the issues (swelling, stability under irradiation, gas retention and release) of using Mo as matrix to transmute Am, two irradiation experiments have been conducted recently at the High Flux Reactor (HFR) in Petten (The Netherland) namely HELIOS and BODEX. The BODEX experiment is a separate effect test, where the molybdenum behaviour is studied without the presence of fission products using 10B to "produce" helium, the HELIOS experiment included a more representative fuel target with the presence of Am and fission product. This paper covers the results of Post Irradiation Examination (PIE) of the two irradiation experiments mentioned above where molybdenum behaviour has been deeply investigated as possible matrix to transmute americium (CerMet fuel target). The behaviour of molybdenum looks satisfying at operating temperature but at high temperature (above 1000 °C) more investigation should be performed.

  9. The behaviour under irradiation of molybdenum matrix for inert matrix fuel containing americium oxide (CerMet concept)

    NASA Astrophysics Data System (ADS)

    D'Agata, E.; Knol, S.; Fedorov, A. V.; Fernandez, A.; Somers, J.; Klaassen, F.

    2015-10-01

    Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors or Accelerator Driven System (ADS, subcritical reactors dedicated to transmutation) of long-lived nuclides like 241Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. In order to safely burn americium in a fast reactor or ADS, it must be incorporated in a matrix that could be metallic (CerMet target) or ceramic (CerCer target). One of the most promising matrix to incorporate Am is molybdenum. In order to address the issues (swelling, stability under irradiation, gas retention and release) of using Mo as matrix to transmute Am, two irradiation experiments have been conducted recently at the High Flux Reactor (HFR) in Petten (The Netherland) namely HELIOS and BODEX. The BODEX experiment is a separate effect test, where the molybdenum behaviour is studied without the presence of fission products using 10B to "produce" helium, the HELIOS experiment included a more representative fuel target with the presence of Am and fission product. This paper covers the results of Post Irradiation Examination (PIE) of the two irradiation experiments mentioned above where molybdenum behaviour has been deeply investigated as possible matrix to transmute americium (CerMet fuel target). The behaviour of molybdenum looks satisfying at operating temperature but at high temperature (above 1000 °C) more investigation should be performed.

  10. Experimental approach and modelling of the mechanical behaviour of graphite fuel elements subjected to compression pulses

    NASA Astrophysics Data System (ADS)

    Forquin, P.

    2010-06-01

    Among the activities led by the Generation IV International Forum (GIF) relative to the future nuclear systems, the improvement of recycling of fuel elements and their components is a major issue. One of the studied systems by the GIF is the graphite-moderated high-temperature gas cooled reactor (HTGR). The fuel elements are composed of fuel roads half-inch in diameter named compacts. The compacts contain spherical particles made of actinide kernels about 500 m in diameter coated with three layers of carbon and silicon carbide, each about 50 m thick, dispersed in a graphite matrix. Recycling of compacts requires first a separation of triso-particles from the graphite matrix and secondly, the separation of the triso-coating from the kernels. This aim may be achieved by using pulsed currents: the compacts are placed within a cell filled by water and exposed to high voltage between 200 - 500 kV and discharge currents from 10 to 20 kA during short laps of time (about 2 µs) [1-2]. This repeated treatment leads to a progressive fragmentation of the graphite matrix and a disassembly of the compacts. In order to improve understanding of the fragmentation properties of compacts a series of quasi-static and dynamic experiments have been conducted with similar cylindrical samples containing 10% (volume fraction) of SiC particles coated in a graphite matrix. First, quasi-static compression tests have been performed to identify the mechanical behaviour of the material at low strain-rates (Fig.1). The experiments reveal a complex elasto-visco-plastic behaviour before a brittle failure. The mechanical response is characterised by a low yield stress (about 1 MPa), a strong strain-hardening in the loading phase and marked hysteresis-loops during unloading-reloading stages. Brittle failure is observed for axial stress about 13 MPa. In parallel, a series of flexural tests have been performed with the aim to characterise the quasi-static tensile strength of the particulate

  11. 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.

  12. 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.

  13. FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Duffy, J.G. Jr.

    1961-05-30

    A lattice type fissionable fuel structure for a nuclear reactor is described. The fissionable material is formed into a plurality of rod-llke bodies with each encased in a fluid-tight jacket. A plurality of spaced longitudinal fins are mounted on the exterior and extend radially from each jacket, with a portion of the fins extending radially beyond the remainder of the fins. A collar of short length for each body is mounted on the extended fins for spacing the bodies, and adjacent bodies abut each other through these collars. Should distortion of the bodies take place, coilapse of the outer fins is limited by the shorter flns, thereby insuring some coolant flow at all times. (AEC)

  14. 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.

  15. 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.

  16. 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.

  17. Fuel Element for a Nuclear Reactor

    DOEpatents

    Duffy, Jr., J. G.

    1961-05-30

    A lattice-type fissionable fuel structure for a nuclear reactor is offered. The fissionable material is formed into a plurality of rod-like bodies each encased in a fluid-tight jacket. A plurality of spaced longitudinal fins are mounted on the exterior of and extend radially from each jacket, and a portion of the fins extends radially beyond the remainder of the fins. A collar of short lengih for each body is mounted on the extended fins for spacing the bodies, and adjacent bodies abut each other through these collars. Should distortion of the bodies take place, collapse of the outer fins is limited by the shorter fins thereby insuring some coolant flow therethrough at all times.

  18. 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.

  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. 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).

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

    DOEpatents

    Roake, W.E.; Evans, E.A.; Brite, D.W.

    1960-06-21

    A method of preparing a fuel element for a nuclear reactor is given in which an internally and externally cooled fuel element consisting of two coaxial tubes having a plurality of integral radial ribs extending between the tubes and containing a powdered fuel material is isostatically pressed to form external coolant channels and compact the powder simultaneously.

  2. 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.

  3. 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.

  4. 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.

  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. 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.

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Behaviour of irradiated PHWR fuel pins during high temperature heating

    NASA Astrophysics Data System (ADS)

    Viswanathan, U. K.; Unnikrishnan, K.; Mishra, Prerna; Banerjee, Suparna; Anantharaman, S.; Sah, D. N.

    2008-12-01

    Fuel pins removed from an irradiated pressurised heavy water reactor (PHWR) fuel bundle discharged after an extended burn up of 15,000 MWd/tU have been subjected to isothermal heating tests in temperature range 700-1300 °C inside hot-cells. The heating of the fuel pins was carried out using a specially designed remotely operable furnace, which allowed localized heating of about 100 mm length of the fuel pin at one end under flowing argon gas or in air atmosphere. Post-test examination performed in the hot-cells included visual examination, leak testing, dimension measurement and optical and scanning electron microscopy. Fuel pins having internal pressure of 2.1-2.7 MPa due to fission gas release underwent ballooning and micro cracking during heating for 10 min at 800 °C and 900 °C but not at 700 °C. Fuel pin heated at 1300 °C showed complete disruption of cladding in heating zone, due to the embrittlement of the cladding. The examination of fuel from the pin tested at 1300 °C showed presence of large number of bubbles; both intragranular as well as intergranular bubbles. Details of the experiments and the results are presented in this paper.

  15. Modelling the behaviour of oxide fuels containing minor actinides with urania, thoria and zirconia matrices in an accelerator-driven system

    NASA Astrophysics Data System (ADS)

    Sobolev, V.; Lemehov, S.; Messaoudi, N.; Van Uffelen, P.; Aı̈t Abderrahim, H.

    2003-06-01

    The Belgian Nuclear Research Centre, SCK • CEN, is currently working on the pre-design of the multipurpose accelerator-driven system (ADS) MYRRHA. A demonstration of the possibility of transmutation of minor actinides and long-lived fission products with a realistic design of experimental fuel targets and prognosis of their behaviour under typical ADS conditions is an important task in the MYRRHA project. In the present article, the irradiation behaviour of three different oxide fuel mixtures, containing americium and plutonium - (Am,Pu,U)O 2- x with urania matrix, (Am,Pu,Th)O 2- x with thoria matrix and (Am,Y,Pu,Zr)O 2- x with inert zirconia matrix stabilised by yttria - were simulated with the new fuel performance code MACROS, which is under development and testing at the SCK • CEN. All the fuel rods were considered to be of the same design and sizes: annular fuel pellets, helium bounded with the stainless steel cladding, and a large gas plenum. The liquid lead-bismuth eutectic was used as coolant. Typical irradiation conditions of the hottest fuel assembly of the MYRRHA subcritical core were pre-calculated with the MCNPX code and used in the following calculations as the input data. The results of prediction of the thermo-mechanical behaviour of the designed rods with the considered fuels during three irradiation cycles of 90 EFPD are presented and discussed.

  16. Spent Nuclear Fuel (SNF) Project Product Specification

    SciTech Connect

    PAJUNEN, A.L.

    2000-12-07

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  17. 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%.

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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)

  6. 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.

  7. 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.

  8. Disposition of ORNL's Spent Nuclear Fuel

    SciTech Connect

    Turner, D. W.; DeMonia, B. C.; Horton, L. L.

    2002-02-26

    This paper describes the process of retrieving, repackaging, and preparing Oak Ridge spent nuclear fuel (SNF) for off-site disposition. The objective of the Oak Ridge SNF Project is to safely, reliably, and efficiently manage SNF that is stored on the Oak Ridge Reservation until it can be shipped off-site. The project required development of several unique processes and the design and fabrication of special equipment to enable the successful retrieval, transfer, and repackaging of Oak Ridge SNF. SNF was retrieved and transferred to a hot cell for repackaging. After retrieval of SNF packages, the storage positions were decontaminated and stainless steel liners were installed to resolve the vulnerability of water infiltration. Each repackaged SNF canister has been transferred from the hot cell back to dry storage until off-site shipments can be made. Three shipments of aluminum-clad SNF were made to the Savannah River Site (SRS), and five shipments of non-aluminum-clad SNF are planned to the Idaho National Engineering and Environmental Laboratory (INEEL). Through the integrated cooperation of several organizations including the U.S. Department of Energy (DOE), Bechtel Jacobs Company LLC (BJC), Oak Ridge National Laboratory (ORNL), and various subcontractors, preparations for the disposition of SNF in Oak Ridge have been performed in a safe and successful manner.

  9. 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.

  10. 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.

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

    DOE PAGESBeta

    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,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

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. Metallic Fast Reactor Fuel Fabrication for Global Nuclear Energy Partnership

    SciTech Connect

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter

    2009-07-01

    Fast reactors are once again being considered for nuclear power generation, in addition to transmutation of long-lived fission products resident in spent nuclear fuels. This re-consideration follows with intense developmental programs for both fuel and reactor design. One of the two leading candidates for next generation fast reactor fuel is metal alloys, resulting primarily from the successes achieved in the 1960s to early 1990s with both the experimental breeding reactor-II and the fast flux test facility. The goal of the current program is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional, fast-spectrum nuclear fuel while destroying recycled actinides, thereby closing the nuclear fuel cycle. In order to meet this goal, the program must develop efficient and safe fuel fabrication processes designed for remote operation. This paper provides an overview of advanced casting processes investigated in the past, and the development of a gaseous diffusion calculation that demonstrates how straightforward process parameter modification can mitigate the loss of volatile minor actinides in the metal alloy melt.

  18. 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.

  19. 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.

  20. Microbiology of spent nuclear fuel storage basins.

    PubMed

    Santo Domingo, J W; Berry, C J; Summer, M; Fliermans, C B

    1998-12-01

    Microbiological studies of spent nuclear fuel storage basins at Savannah River Site (SRS) were performed as a preliminary step to elucidate the potential for microbial-influenced corrosion (MIC) in these facilities. Total direct counts and culturable counts performed during a 2-year period indicated microbial densities of 10(4) to 10(7) cells/ml in water samples and on submerged metal coupons collected from these basins. Bacterial communities present in the basin transformed between 15% and 89% of the compounds present in Biologtrade mark plates. Additionally, the presence of several biocorrosion-relevant microbial groups (i.e., sulfate-reducing bacteria and acid-producing bacteria) was detected with commercially available test kits. Scanning electron microscopy and X-ray spectra analysis of osmium tetroxide-stained coupons demonstrated the development of microbial biofilm communities on some metal coupons submerged for 3 weeks in storage basins. After 12 months, coupons were fully covered by biofilms, with some deterioration of the coupon surface evident at the microscopical level. These results suggest that, despite the oligotrophic and radiological environment of the SRS storage basins and the active water deionization treatments commonly applied to prevent electrochemical corrosion in these facilities, these conditions do not prevent microbial colonization and survival. Such microbial densities and wide diversity of carbon source utilization reflect the ability of the microbial populations to adapt to these environments. The presumptive presence of sulfate-reducing bacteria and acid-producing bacteria and the development of biofilms on submerged coupons indicated that an environment for MIC of metal components in the storage basins may occur. However, to date, there has been no indication or evidence of MIC in the basins. Basin chemistry control and corrosion surveillance programs instituted several years ago have substantially abated all corrosion mechanisms

  1. Spent nuclear fuel assembly inspection using neutron computed tomography

    NASA Astrophysics Data System (ADS)

    Pope, Chad Lee

    The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Estimating Source Terms for Diverse Spent Nuclear Fuel Types

    SciTech Connect

    Brett Carlsen; Layne Pincock

    2004-11-01

    The U.S. Department of Energy (DOE) National Spent Nuclear Fuel Program is responsible for developing a defensible methodology for determining the radionuclide inventory for the DOE spent nuclear fuel (SNF) to be dispositioned at the proposed Monitored Geologic Repository at the Yucca Mountain Site. SNF owned by DOE includes diverse fuels from various experimental, research, and production reactors. These fuels currently reside at several DOE sites, universities, and foreign research reactor sites. Safe storage, transportation, and ultimate disposal of these fuels will require radiological source terms as inputs to safety analyses that support design and licensing of the necessary equipment and facilities. This paper summarizes the methodology developed for estimating radionuclide inventories associated with DOE-owned SNF. The results will support development of design and administrative controls to manage radiological risks and may later be used to demonstrate conformance with repository acceptance criteria.

  19. Fission product partitioning in aerosol release from simulated spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Di Lemma, F. G.; Colle, J. Y.; Rasmussen, G.; Konings, R. J. M.

    2015-10-01

    Aerosols created by the vaporization of simulated spent nuclear fuel (simfuel) were produced by laser heating techniques and characterised by a wide range of post-analyses. In particular attention has been focused on determining the fission product behaviour in the aerosols, in order to improve the evaluation of the source term and consequently the risk associated with release from spent fuel sabotage or accidents. Different simulated spent fuels were tested with burn-up up to 8 at. %. The results from the aerosol characterisation were compared with studies of the vaporization process by Knudsen Effusion Mass Spectrometry and thermochemical equilibrium calculations. These studies permit an understanding of the aerosol gaseous precursors and the gaseous reactions taking place during the aerosol formation process.

  20. 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.

  1. 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.

  2. K-Basin spent nuclear fuel characterization data report

    SciTech Connect

    Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

    1995-11-01

    The spent nuclear fuel (SNF) project characterization activities will be furnishing technical data on SNF stored at the K Basins in support of a pathway for placement of a ``stabilized`` form of SNF into an interim storage facility. This report summarizes the results so far of visual inspection of the fuel samples, physical characterization (e.g., weight and immersion density measurements), metallographic examinations, and controlled atmosphere furnace testing of three fuel samples shipped from the KW Basin to the Postirradiation Testing Laboratory (PTL). Data on sludge material collected by filtering the single fuel element canister (SFEC) water are also discussed in this report.

  3. 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.

  4. 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.

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

    DOEpatents

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

    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.

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

    SciTech Connect

    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.

  7. 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.

  8. 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

  9. Understanding electrostatic charge behaviour in aircraft fuel systems

    NASA Astrophysics Data System (ADS)

    Ogilvy, Jill A.; Hooker, Phil; Bennett, Darrell

    2015-10-01

    This paper presents work on the simulation of electrostatic charge build-up and decay in aircraft fuel systems. A model (EC-Flow) has been developed by BAE Systems under contract to Airbus, to allow the user to assess the effects of changes in design or in refuel conditions. Some of the principles behind the model are outlined. The model allows for a range of system components, including metallic and non-metallic pipes, valves, filters, junctions, bends and orifices. A purpose-built experimental rig was built at the Health and Safety Laboratory in Buxton, UK, to provide comparison data. The rig comprises a fuel delivery system, a test section where different components may be introduced into the system, and a Faraday Pail for measuring generated charge. Diagnostics include wall currents, charge densities and pressure losses. This paper shows sample results from the fitting of model predictions to measurement data and shows how analysis may be used to explain some of the observed trends.

  10. 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.