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Sample records for adjacent fuel elements

  1. 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element, when separated from other fuel elements, fits loosely in the coolant tube and so can easily be inserted or withdrawn from the tube.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)

  2. FUEL ELEMENT

    DOEpatents

    Bean, R.W.

    1963-11-19

    A ceramic fuel element for a nuclear reactor that has improved structural stability as well as improved cooling and fission product retention characteristics is presented. The fuel element includes a plurality of stacked hollow ceramic moderator blocks arranged along a tubular raetallic shroud that encloses a series of axially apertured moderator cylinders spaced inwardly of the shroud. A plurality of ceramic nuclear fuel rods are arranged in the annular space between the shroud and cylinders of moderator and appropriate support means and means for directing gas coolant through the annular space are also provided. (AEC)

  3. FUEL ELEMENT

    DOEpatents

    Fortescue, P.; Zumwalt, L.R.

    1961-11-28

    A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

  4. FUEL ELEMENT

    DOEpatents

    Howard, R.C.; Bokros, J.C.

    1962-03-01

    A fueled matrlx eontnwinlng uncomblned carbon is deslgned for use in graphlte-moderated gas-cooled reactors designed for operatlon at temperatures (about 1500 deg F) at which conventional metallic cladding would ordlnarily undergo undesired carburization or physical degeneratlon. - The invention comprlses, broadly a fuel body containlng uncombined earbon, clad with a nickel alloy contalning over about 28 percent by' weight copper in the preferred embodlment. Thls element ls supporirted in the passageways in close tolerance with the walls of unclad graphite moderator materlal. (AEC)

  5. 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)

  6. FUEL ELEMENT SUPPORT

    DOEpatents

    Wyman, W.L.

    1961-06-27

    The described cylindrical fuel element has longitudinally spaced sets of short longitudinal ribs circumferentially spaced from one another. The ribs support the fuel element in a coolant tube so that there is an annular space for coolant flow between the fuel element and the interior of the coolant tube. If the fuel element grows as a result of reactor operation, the circumferential distribution of the ribs maintains the uniformity of the annular space between the coolant tube and the fuel element, and the collapsibility of the ribs prevents the fuel element from becoming jammed in the coolant tube.

  7. COMPARTMENTED REACTOR FUEL ELEMENT

    DOEpatents

    Cain, F.M. Jr.

    1962-09-11

    A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

  8. 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.

  9. 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.

  10. Neutronic fuel element fabrication

    DOEpatents

    Korton, George

    2004-02-24

    This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure

  11. COMPOSITE FUEL ELEMENT

    DOEpatents

    Hurford, W.J.; Gordon, R.B.; Johnson, W.A.

    1962-12-25

    A sandwich-type fuel element for a reactor is described. This fuel element has the shape of an elongated flat plate and includes a filler plate having a plurality of compartments therein in which the fuel material is located. The filler plate is clad on both sides with a thin cladding material which is secured to the filler plate only to completely enclose the fuel material in each compartment. (AEC)

  12. 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.

  13. 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)

  14. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Shackleford, M.H.

    1958-12-16

    A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

  15. 12. LOG FOUNDATION ELEMENTS OF THE SAWMILL ADJACENT TO THE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    12. LOG FOUNDATION ELEMENTS OF THE SAWMILL ADJACENT TO THE CANAL, LOOKING EAST. BARREN AREA IN FOREGROUND IS DECOMPOSING SAWDUST. DIRT PILE IN BACKGROUND IS THE EDGE OF THE SUMMIT COUNTY LANDFILL. - Snake River Ditch, Headgate on north bank of Snake River, Dillon, Summit County, CO

  16. JACKETED FUEL ELEMENT

    DOEpatents

    Wigner, E.P.; Szilard, L.; Creutz, E.C.

    1959-02-01

    These fuel elements are comprised of a homogeneous metallic uranium body completely enclosed and sealed in an aluminum cover. The uranium body and aluminum cover are bonded together by a layer of zinc located between them. The bonding layer serves to improve transfer of heat, provides an additional protection against corrosion of the uranium by the coolant, and also localizes any possible corrosion by preventing travel of corrosive material along the surface of the fuel element.

  17. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Gurinsky, D.H.; Powell, R.W.; Fox, M.

    1959-11-24

    A nuclear fuel element comprising a plurality of nuclear fuel bearing strips is presented. The strips are folded along their longitudinal axes to an angle of about 60 deg and are secured at each end by ferrule to form an elongated assembly suitable for occupying a cylindrical coolant channel.

  18. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Kesselring, K.A.; Seybolt, A.U.

    1958-12-01

    A reactor fuel element of the capillary tube type is described. The element consists of a thin walled tube, sealed at both ends, and having an interior coatlng of a fissionable material, such as uranium enriched in U-235. The tube wall is gas tight and is constructed of titanium, zirconium, or molybdenum.

  19. 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.

  20. FUEL ELEMENT CONSTRUCTION

    DOEpatents

    Simnad, M.T.

    1961-08-15

    A method of preventing diffusible and volatile fission products from diffusing through a fuel element container and contaminating reactor coolant is described. More specifically, relatively volatile and diffusible fission products either are adsorbed by or react with magnesium fluoride or difluoride to form stable, less volatile, less diffusible forms. The magnesium fluoride or difluoride is disposed anywhere inwardly from the outer surface of the fuel element container in order to be contacted by the fission products before they reach and contaminate the reactor coolant. (AEC)

  1. FUEL ELEMENT CONSTRUCTION

    DOEpatents

    Zumwalt, L.R.

    1961-08-01

    Fuel elements having a solid core of fissionable material encased in a cladding material are described. A conversion material is provided within the cladding to react with the fission products to form stable, relatively non- volatile compounds thereby minimizing the migration of the fission products into the coolant. The conversion material is preferably a metallic fluoride, such as lead difluoride, and may be in the form of a coating on the fuel core or interior of the cladding, or dispersed within the fuel core. (AEC)

  2. CONCENTRIC TUBE FUEL ELEMENT SPRING ALIGNMENT SPACER DEVICE

    DOEpatents

    Weems, S.J.

    1963-09-24

    A rib construction for a nuclear-fuel element is described, in which one of three peripherally spaced ribs adjacent to each end of the fuel element is mounted on a radially yielding spring that embraces the fuel element. This spring enables the fuel element to have a good fit with a coolant tube and yet to be easily inserted in and withdrawn from the tube. (AEC)

  3. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Stacy, J.T.

    1958-12-01

    A reactor fuel element having a core of molybdenum-uranium alloy jacketed in stainless steel is described. A barrier layer of tungsten, tantalum, molybdenum, columbium, or silver is interposed between the core and jacket to prevent formation of a low melting eutectic between uranium and the varlous alloy constituents of the stainless steel.

  4. JACKETED REACTOR FUEL ELEMENT

    DOEpatents

    Smith, K.F.; Van Thyne, R.J.

    1958-12-01

    A fuel element is described for fast reactors comprised of a core of uranium metal containing material and a jacket around the core, the jacket consisting of from 2.5 to 15 percent of titanium, from 1 to 5 percent of niobium, and from 80 to 96.5 percent of vanadium.

  5. 20. Interior view of fuel storage pit or vault adjacent ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    20. Interior view of fuel storage pit or vault adjacent to Test Cell 9 in Component Test Laboratory (T-27), looking west. Photograph shows upgraded instrumentation, piping, tanks, and technological modifications installed in 1997-99 to accommodate component testing requirements for the Atlas V missile. - Air Force Plant PJKS, Systems Integration Laboratory, Components Test Laboratory, Waterton Canyon Road & Colorado Highway 121, Lakewood, Jefferson County, CO

  6. 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.

  7. FUEL ELEMENT FABRICATION METHOD

    DOEpatents

    Hix, J.N.; Cooley, G.E.; Cunningham, J.E.

    1960-05-31

    A method is given for assembling and fabricating a fuel element comprising a plurality of spaced parallel fuel plates of a bowed configuration supported by and between a pair of transperse aluminum side plates. In this method, a brasing alloy is preplated on one surface of the aluminum side plates in the form of a cladding or layer-of uniform thickness. Grooves are then cut into the side plates through the alloy layer and into the base aluminum which results in the utilization of thinner aluminum side plates since a portion of the necessary groove depth is supplied by the brazing alloy.

  8. 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)

  9. 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.

  10. 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.

  11. RECONDITIONING FUEL ELEMENTS

    DOEpatents

    Brandt, H.L.

    1962-02-20

    A process is given for decanning fuel elements that consist of a uranium core, an intermediate section either of bronze, silicon, Al-Si, and uranium silicide layers or of lead, Al-Si, and uranium silicide layers around said core, and an aluminum can bonded to said intermediate section. The aluminum can is dissolved in a solution of sodium hydroxide (9 to 20 wt%) and sodium nitrate (35 to 12 wt %), and the layers of the intermediate section are dissolved in a boiling sodium hydroxide solution of a minimum concentration of 50 wt%. (AEC) A method of selectively reducing plutonium oxides and the rare earth oxides but not uranium oxides is described which comprises placing the oxides in a molten solvent of zinc or cadmium and then adding metallic uranium as a reducing agent. (AEC)

  12. 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.

  13. 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.

  14. 15. VIEW OF DUMMY FUEL ELEMENT ON FUEL ELEMENT HOLDER. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    15. VIEW OF DUMMY FUEL ELEMENT ON FUEL ELEMENT HOLDER. SHOWS AIR FORCE MAN AT EDGE OF TANK. INEL PHOTO NUMBER 65-6176, TAKEN NOVEMBER 10, 1965. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

  15. 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)

  16. 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.

  17. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-11-21

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

  18. REACTOR FUEL ELEMENTS TESTING CONTAINER

    DOEpatents

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  19. Spent graphite fuel element processing

    SciTech Connect

    Holder, N.D.; Olsen, C.W.

    1981-07-01

    The Department of Energy currently sponsors two programs to demonstrate the processing of spent graphite fuel elements. General Atomic in San Diego operates a cold pilot plant to demonstrate the processing of both US and German high-temperature reactor fuel. Exxon Nuclear Idaho Company is demonstrating the processing of spent graphite fuel elements from Rover reactors operated for the Nuclear Rocket Propulsion Program. This work is done at Idaho National Engineering Laboratory, where a hot facility is being constructed to complete processing of the Rover fuel. This paper focuses on the graphite combustion process common to both programs.

  20. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Horning, W.A.; Lanning, D.D.; Donahue, D.J.

    1959-10-01

    A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

  1. 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)

  2. FUEL ELEMENT FOR NEUTRONIC REACTORS

    DOEpatents

    Evans, T.C.; Beasley, E.G.

    1961-01-17

    A fuel element for neutronic reactors, particularly the gas-cooled type of reactor, is described. The element comprises a fuel-bearing plate rolled to form a cylinder having a spiral passageway passing from its periphery to its center. In operation a coolant is admitted to the passageway at the periphery of the element, is passed through the spiral passageway, and emerges into a central channel defined by the inner turn of the rolled plate. The advantage of the element is that the fully heated coolant (i.e., coolant emerging into the central channel) is separated and thus insulated from the periphery of the element, which may be in contact with a low-temperature moderator, by the intermediate turns of the spiral fuel element.

  3. Cleanup Verification Package for the 118-F-8:4 Fuel Storage Basin West Side Adjacent and Side Slope Soils

    SciTech Connect

    L. D. Habel

    2008-03-18

    This cleanup verification package documents completion of remedial action, sampling activities, and compliance with cleanup criteria for the 118-F-8:4 Fuel Storage Basin West Side Adjacent and Side Slope Soils. The rectangular-shaped concrete basin on the south side of the 105-F Reactor building served as an underwater collection, storage, and transfer facility for irradiated fuel elements discharged from the reactor.

  4. FUEL ELEMENTS FOR NEUTRONIC REACTORS

    DOEpatents

    Foote, F.G.; Jette, E.R.

    1963-05-01

    A fuel element for a nuclear reactor is described that consists of a jacket containing a unitary core of fissionable material and a filling of a metal of the group consisting of sodium and sodium-potassium alloys. (AEC)

  5. Identification of leaking TRIGA fuel elements

    SciTech Connect

    Bennion, John S.; Crawford, Kevan C.; Gansauge, Todd C.; Sandquist, Gary M.

    1990-07-01

    -steel-clad fuel elements were eventually swapped. After exchanging nearly one-half of the incore fuel elements without any noticeable decrease in leaked fission-product activity, it was apparent that multiple defective elements were probably responsible for the leakage. A more aggressive identification program was implemented whereby water samples were taken directly above each fuel element in the core during operation at 90 kW. A small weighted polyethylene funnel was connected to 1/8-inch Tygon tubing extending from the reactor pool to an adjacent laboratory containing gamma spectroscopy equipment. The tubing was wrapped around a high-purity germanium detector and coupled to a small peristaltic pump adjusted to provide a flow velocity of approximately one foot-per-second and drained into a large bucket. The detector signal was fed into a multichannel scaler calibrated to measure the 151 keV photon emitted by Kr-85m. The funnel was manually suspended above each core fuel position for one minute sampling water convected along each fuel element. A careful record of the time during which each position was sampled was maintained so that elements with high krypton activity could be precisely located. After three such operations, a total of seven stainless-steel elements were identified as candidates contributing to the leakage and were replaced. Subsequent operations have shown no fission-product release and confirmed that the fuel responsible for the leakage had been removed from the core. Evaluation of the candidate leaking elements is planned to identify actual 'leakers' in the future. (author)

  6. IMPROVED TYPE OF FUEL ELEMENT

    DOEpatents

    Monson, H.O.

    1961-01-24

    A radiator-type fuel block assembly is described. It has a hexagonal body of neutron fissionable material having a plurality of longitudinal equal- spaced coolant channels therein aligned in rows parallel to each face of the hexagonal body. Each of these coolant channels is hexagonally shaped with the corners rounded and enlarged and the assembly has a maximum temperature isothermal line around each channel which is approximately straight and equidistant between adjacent channels.

  7. 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.

  8. 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)

  9. Compact Fuel Element Environment Test

    NASA Technical Reports Server (NTRS)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

    2012-01-01

    Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

  10. 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.

  11. Low cost, lightweight fuel cell elements

    NASA Technical Reports Server (NTRS)

    Kindler, Andrew (Inventor)

    2001-01-01

    New fuel cell elements for use in liquid feed fuel cells are provided. The elements including biplates and endplates are low in cost, light in weight, and allow high efficiency operation. Electrically conductive elements are also a part of the fuel cell elements.

  12. High performance fuel element with end seal

    DOEpatents

    Lee, Gary E.; Zogg, Gordon J.

    1987-01-01

    A nuclear fuel element comprising an elongate block of refractory material having a generally regular polygonal cross section. The block includes parallel, spaced, first and second end surfaces. The first end surface has a peripheral sealing flange formed thereon while the second end surface has a peripheral sealing recess sized to receive the flange. A plurality of longitudinal first coolant passages are positioned inwardly of the flange and recess. Elongate fuel holes are separate from the coolant passages and disposed inwardly of the flange and the recess. The block is further provided with a plurality of peripheral second coolant passages in general alignment with the flange and the recess for flowing coolant. The block also includes two bypasses for each second passage. One bypass intersects the second passage adjacent to but spaced from the first end surface and intersects a first passage, while the other bypass intersects the second passage adjacent to but spaced from the second end surface and intersects a first passage so that coolant flowing through the second passages enters and exits the block through the associated first passages.

  13. Fuel elements of thermionic converters

    SciTech Connect

    Hunter, R.L.; Gontar, A.S.; Nelidov, M.V.; Nikolaev, Yu.V.; Schulepov, L.N.

    1997-01-01

    Work on thermionic nuclear power systems has been performed in Russia within the framework of the TOPAZ reactor program since the early 1960s. In the TOPAZ in-core thermionic convertor reactor design, the fuel element`s cladding is also the thermionic convertor`s emitter. Deformation of the emitter can lead to short-circuiting and is the primary cause of premature TRC failure. Such deformation can be the result of fuel swelling, thermocycling, or increased unilateral pressure on the emitter due to the release of gaseous fission products. Much of the work on TRCs has concentrated on preventing or mitigating emitter deformation by improving the following materials and structures: nuclear fuel; emitter materials; electrical insulators; moderator and reflector materials; and gas-exhaust device. In addition, considerable effort has been directed toward the development of experimental techniques that accurately mimic operational conditions and toward the creation of analytical and numerical models that allow operational conditions and behavior to be predicted without the expense and time demands of in-pile tests. New and modified materials and structures for the cores of thermionic NPSs and new fabrication processes for the materials have ensured the possibility of creating thermionic NPSs for a wide range of powers, from tens to several hundreds of kilowatts, with life spans of 5 to 10 years.

  14. Stress Reduction in Adjacent Level Discs via Dynamic Instrumentation: A Finite Element Analysis

    PubMed Central

    Castellvi, Antonio E.; Huang, Hao; Vestgaarden, Tov; Saigal, Sunil; Pienkowski, David

    2007-01-01

    Background Conventional (rigid) fusion instrumentation is believed to accelerate the degeneration of adjacent discs by increasing stresses caused by motion discontinuity. Fusion instrumentation that employs reduced rod stiffness and increased axial motion, or dynamic instrumentation, may partially alleviate this problem, but the effects of this instrumentation on the stresses in the adjacent disc are unknown. We used a finiteelement model to calculate and compare the stresses in the adjacent-level disc that are induced by rigid and dynamic posterior lumbar fusion instrumentation. Methods A 3-dimensional finite-element model of the lumbar spine was obtained that simulated flexion and extension. The L5–S1 segment of this model was fused, and the L4–L5 segment was fixed with rigid or dynamic instrumentation. The mechanical properties of the dynamic instrumentation were determined by laboratory testing and then used in the finite-element model. Peak stresses in the lumbar discs were calculated and compared. Results The reduced-stiffness component of the dynamic instrumentation was associated with a 1% to 2% reduction in peak compressive stresses in the adjacent-level disc (at 45° flexion), and the increased axial motion component of this instrumentation reduced peak disc stress by 8% to 9%. Areas of disc tissue exposed to 80% of peak stresses of 6.17 MPa were 47% less for discs adjacent to dynamic instrumentation than for those adjacent to rigid instrumentation. Conclusions Reduced stiffness and increased axial motion of dynamic posterior lumbar fusion instrumentation designs result in an approximately 10% cumulative stress reduction for each flexion cycle. The effect of this stress reduction over many cycles may be substantial. Clinical Relevance The cumulative effect of this reduced amplitude and distribution of peak stresses in the adjacent disc may partially alleviate the problem of adjacent-level disc degeneration. PMID:25802582

  15. METHOD OF MAKING FUEL ELEMENTS

    DOEpatents

    Bean, C.H.; Macherey, R.E.

    1959-12-01

    A method is described for fabricating fuel elements, particularly for enclosing a plate of metal with a second metal by inserting the plate into an aperture of a frame of a second plate, placing a sheet of the second metal on each of opposite faces of the assembled plate and frame, purging with an inert gas the air from the space within the frame and the sheets while sealing the seams between the frame and the sheets, exhausting the space, purging the space with air, re-exhausting the spaces, sealing the second aperture, and applying heat and pressure to bond the sheets, the plate, and the frame to one another.

  16. Finite element analysis of advanced neutron source fuel plates

    SciTech Connect

    Luttrell, C.R.

    1995-08-01

    The proposed design for the Advanced Neutron Source reactor core consists of closely spaced involute fuel plates. Coolant flows between the plates at high velocities. It is vital that adjacent plates do not come in contact and that the coolant channels between the plates remain open. Several scenarios that could result in problems with the fuel plates are studied. Finite element analyses are performed on fuel plates under pressure from the coolant flowing between the plates at a high velocity, under pressure because of a partial flow blockage in one of the channels, and with different temperature profiles.

  17. Pounding Effects on the Earthquake Response of Adjacent Reinforced Concrete Structures Strengthened by Cable Elements

    NASA Astrophysics Data System (ADS)

    Liolios, Angelos; Liolios, Asterios; Hatzigeorgiou, George; Radev, Stefan

    2014-06-01

    A numerical approach for estimating the effects of pounding (seismic interaction) on the response of adjacent Civil Engineering structures is presented. Emphasis is given to reinforced concrete (RC) frames of existing buildings which are seismically strengthened by cable-elements. A double discretization, in space by the Finite Element Method and in time by a direct incremental approach is used. The unilateral behaviours of both, the cable-elements and the interfaces contact-constraints, are taken strictly into account and result to inequality constitutive conditions. So, in each time-step, a non-convex linear complementarity problem is solved. It is found that pounding and cable strengthening have significant effects on the earthquake response and, hence, on the seismic upgrading of existing adjacent RC structures.

  18. 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.

  19. Sacroiliac Joint Fusion Minimally Affects Adjacent Lumbar Segment Motion: A Finite Element Study

    PubMed Central

    Kiapour, Ali; Yerby, Scott A.; Goel, Vijay K.

    2015-01-01

    Background Adjacent segment disease is a recognized consequence of fusion in the spinal column. Fusion of the sacroiliac joint is an effective method of pain reduction. Although effective, the consequences of sacroiliac joint fusion and the potential for adjacent segment disease for the adjacent lumbar spinal levels is unknown. The objective of this study was to quantify the change in range of motion of the sacroiliac joint and the adjacent lumbar spinal motion segments due to sacroiliac joint fusion and compare these changes to previous literature to assess the potential for adjacent segment disease in the lumbar spine. Methods An experimentally validated finite element model of the lumbar spine and pelvis was used to simulate a fusion of the sacroiliac joint using three laterally placed triangular implants (iFuse Implant System, SI-BONE, Inc., San Jose, CA). The range of motion of the sacroiliac joint and the adjacent lumbar spinal motion segments were calculated using a hybrid loading protocol and compared with the intact range of motion in flexion, extension, lateral bending, and axial rotation. Results The range of motions of the treated sacroiliac joints were reduced in flexion, extension, lateral bending, and axial rotation, by 56.6%, 59.5%, 27.8%, and 53.3%, respectively when compared with the intact condition. The stiffening of the sacroiliac joint resulted in increases at the adjacent lumbar motion segment (L5-S1) for flexion, extension, lateral bending, and axial rotation, of 3.0%, 3.7%, 1.1%, and 4.6%, respectively. Conclusions Fusion of the sacroiliac joint resulted in substantial (> 50%) reductions in flexion, extension, and axial rotation of the sacroiliac joint with minimal (< 5%) increases in range of motion in the lumbar spine. Although the predicted increases in lumbar range of motion are minimal after sacroiliac joint fusion, the long-term clinical results remain to be investigated. PMID:26767156

  20. Visual examinations of K east fuel elements

    SciTech Connect

    Pitner, A.L., Fluor Daniel Hanford

    1997-02-03

    Selected fuel elements stored in both ``good fuel`` and ``bad fuel`` canisters in K East Basin were extracted and visually examined full length for damage. Lower end damage in the ``bad fuel`` canisters was found to be more severe than expected based on top end appearances. Lower end damage for the ``good fuel`` canisters, however, was less than expected based on top end observations. Since about half of the fuel in K East Basin is contained in ``good fuel`` canisters based on top end assessments, the fraction of fuel projected to be intact with respect to IPS processing considerations remains at 50% based on these examination results.

  1. DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

    DOEpatents

    Horn, F.L.

    1961-12-12

    Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

  2. 35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ELEMENT HOLDER, TRIP MECHANISM COVER, AND OTHER DETAILS. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-3. INEL INDEX CODE NUMBER: 075 0701 60 851 151977. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

  3. NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

    DOEpatents

    Szilard, L.; Young, G.J.

    1958-03-01

    This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

  4. Minor and trace element concentrations in adjacent kamacite and taenite in the Krymka chondrite

    NASA Astrophysics Data System (ADS)

    Meftah, N.; Mostefaoui, S.; Jambon, A.; Guedda, E. H.; Pont, S.

    2016-04-01

    We report in situ NanoSIMS siderophile minor and trace element abundances in individual Fe-Ni metal grains in the unequilibrated chondrite Krymka (LL3.2). Associated kamacite and taenite of 10 metal grains in four chondrules and one matrix metal were analyzed for elemental concentrations of Fe, Ni, Co, Cu, Rh, Ir, and Pt. The results show large elemental variations among the metal grains. However, complementary and correlative variations exist between adjacent kamacite-taenite. This is consistent with the unequilibrated character of the chondrite and corroborates an attainment of chemical equilibrium between the metal phases. The calculated equilibrium temperature is 446 ± 9 °C. This is concordant with the range given by Kimura et al. (2008) for the Krymka postaccretion thermal metamorphism. Based on Ni diffusivity in taenite, a slow cooling rate is estimated of the Krymka parent body that does not exceed ~1K Myr-1, which is consistent with cooling rates inferred by other workers for unequilibrated ordinary chondrites. Elemental ionic radii might have played a role in controlling elemental partitioning between kamacite and taenite. The bulk compositions of the Krymka metal grains have nonsolar (mostly subsolar) element/Ni ratios suggesting the Fe-Ni grains could have formed from distinct precursors of nonsolar compositions or had their compositions modified subsequent to chondrule formation events.

  5. MRT fuel element inspection at Dounreay

    SciTech Connect

    Gibson, J.

    1997-08-01

    To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

  6. Identification of failed fuel element

    DOEpatents

    Fryer, Richard M.; Matlock, Robert G.

    1976-06-22

    A passive fission product gas trap is provided in the upper portion of each fuel subassembly in a nuclear reactor. The gas trap consists of an inverted funnel of less diameter than the subassembly having a valve at the apex thereof. An actuating rod extends upwardly from the valve through the subassembly to a point where it can be contacted by the fuel handling mechanism for the reactor. Interrogation of the subassembly for the presence of fission products is accomplished by lowering the fuel handling machine onto the subassembly to press down on the actuating rod and open the valve.

  7. Apparatus for inspecting fuel elements

    DOEpatents

    Oakley, David J.; Groves, Oliver J.; Kaiser, Bruce J.

    1986-01-01

    Disclosed is an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes. After inspection, the fuel pins are lowered by the pin lifting and electrode operating means into a second roller assembly which longitudinally conveys approved pins from the airtight enclosure in which the alpha monitor is mounted. If the fuel pins are rejected then they are moved laterally by a second transfer means and onto another system for further processing.

  8. Apparatus for inspecting fuel elements

    DOEpatents

    Kaiser, B.J.; Oakley, D.J.; Groves, O.J.

    1984-12-21

    This disclosure describes an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes. After inspection, the fuel pins are lowered by the pin lifting and electrode operating means into a second roller assembly which longitudinally conveys approved pins from the airtight enclosure in which the alpha monitor is mounted. If the fuel pins are rejected then they are moved laterally by a second transfer means and onto another system for further processing.

  9. Volume reduction of spent fuel elements for direct disposal

    SciTech Connect

    Wasserfuhr, I.C.

    1995-12-31

    The method of direct disposal of spent fuel elements provides the placing of fuel and non-fuel elements into the POLLUX final disposal casks. It is, however, necessary to disassemble the fuel elements into fuel rods and structural parts. While the fuel rods are condensed, the remaining structure is treated further with a 500-t skeleton press to minimize the volume.

  10. Fuel elements of research reactor CM

    SciTech Connect

    Kozlov, A.V.; Morozov, A.V.; Vatulin, A.V.; Ershov, S.A.

    2013-07-01

    In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

  11. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOEpatents

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  12. 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.

  13. Gas Turbine Engine Staged Fuel Injection Using Adjacent Bluff Body and Swirler Fuel Injectors

    NASA Technical Reports Server (NTRS)

    Snyder, Timothy S. (Inventor)

    2015-01-01

    A fuel injection array for a gas turbine engine includes a plurality of bluff body injectors and a plurality of swirler injectors. A control operates the plurality of bluff body injectors and swirler injectors such that bluff body injectors are utilized without all of the swirler injectors at least at low power operation. The swirler injectors are utilized at higher power operation.

  14. Trace element fingerprinting of cockle (Cerastoderma edule) shells can reveal harvesting location in adjacent areas.

    PubMed

    Ricardo, Fernando; Génio, Luciana; Costa Leal, Miguel; Albuquerque, Rui; Queiroga, Henrique; Rosa, Rui; Calado, Ricardo

    2015-01-01

    Determining seafood geographic origin is critical for controlling its quality and safeguarding the interest of consumers. Here, we use trace element fingerprinting (TEF) of bivalve shells to discriminate the geographic origin of specimens. Barium (Ba), manganese (Mn), magnesium (Mg), strontium (Sr) and lead (Pb) were quantified in cockle shells (Cerastoderma edule) captured with two fishing methods (by hand and by hand-raking) and from five adjacent fishing locations within an estuarine system (Ria de Aveiro, Portugal). Results suggest no differences in TEF of cockle shells captured by hand or by hand-raking, thus confirming that metal rakes do not act as a potential source of metal contamination that could somehow bias TEF results. In contrast, significant differences were recorded among locations for all trace elements analysed. A Canonical Analysis of Principal Coordinates (CAP) revealed that 92% of the samples could be successfully classified according to their fishing location using TEF. We show that TEF can be an accurate, fast and reliable method to determine the geographic origin of bivalves, even among locations separated less than 1 km apart within the same estuarine system. Nonetheless, follow up studies are needed to determine if TEF can reliably discriminate between bivalves originating from different ecosystems. PMID:26149418

  15. Upgraded HFIR Fuel Element Welding System

    SciTech Connect

    Sease, John D

    2010-02-01

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.

  16. 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.

  17. 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.

  18. 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.

  19. JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

    DOEpatents

    Szilard, L.; Wigner, E.P.; Creutz, E.C.

    1959-05-12

    Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

  20. Environmental significance of biogenic elements in surface sediments of the Changjiang Estuary and its adjacent areas.

    PubMed

    Yu, Yu; Song, Jinming; Li, Xuegang; Yuan, Huamao; Li, Ning; Duan, Liqin

    2013-11-01

    Biogenic elements and six phosphorus (P) fractions in surface sediments from the Changjiang Estuary and adjacent waters were determined to investigate the governing factors of these elements, and further to discuss their potential uses as paleo-environment proxies and risks of P release from sediment. Total organic carbon (TOC) and leachable organic P (Lea-OP) showed high concentrations in the estuary, Zhejiang coast and offshore upwelling area. They came from both the Changjiang River and marine biological input. Biogenic silicon (BSi) exhibited a high concentration band between 123 and 124 degree E. BSi mainly came from diatom production and its concentration in the inshore area was diluted by river sediment. Total nitrogen (TN) was primarily of marine biogenic origin. Seaward decreasing trends of Fe-bound P and Al-bound P revealed their terrestrial origins. Influenced by old Huanghe sediment delivered by the Jiangsu coastal current, the maximum concentration of detrital P (Det-P) was observed in the area north of the estuary. Similar high concentrations of carbonate fluorapatite (CFA-P) and CaCO3in the southern study area suggested marine calcium-organism sources of CFA-P. TOC, TN and non-apatite P were enriched in fine sediment, and Det-P partially exhibited coarse-grain enrichment, but BSi had no correlation with sediment grain size. Different sources and governing factors made biogenic elements and P species have distinct potential uses in indicating environmental conditions. Transferable P accounted for 14%-46% of total P. In an aerobic environment, there was low risk of P release from sediment, attributed to excess Fe oxides in sediments. PMID:24552046

  1. FUEL ELEMENT AND METHOD OF PREPARATION

    DOEpatents

    Kingston, W.E.

    1961-04-25

    A nuclear fuel element in the form of a wire is reported. A bar of uranium is enclosed in a thin layer of aluminum and the composite is sheathed in beryllium, zirconium, or stainnless steel. The sheathed article is then drawn to wire form, heated to alloy the aluminum with both uranium and sheath, and finally cold worked.

  2. Liquid fuel injection elements for rocket engines

    NASA Technical Reports Server (NTRS)

    Cox, George B., Jr. (Inventor)

    1993-01-01

    Thrust chambers for liquid propellant rocket engines include three principal components. One of these components is an injector which contains a plurality of injection elements to meter the flow of propellants at a predetermined rate, and fuel to oxidizer mixture ratio, to introduce the mixture into the combustion chamber, and to cause them to be atomized within the combustion chamber so that even combustion takes place. Evolving from these injectors are tube injectors. These tube injectors have injection elements for injecting the oxidizer into the combustion chamber. The oxidizer and fuel must be metered at predetermined rates and mixture ratios in order to mix them within the combustion chamber so that combustion takes place smoothly and completely. Hence tube injectors are subject to improvement. An injection element for a liquid propellant rocket engine of the bipropellant type is provided which includes tangential fuel metering orifices, and a plurality of oxidizer tube injection elements whose injection tubes are also provided with tangential oxidizer entry slots and internal reed valves.

  3. Automatic inspection for remotely manufactured fuel elements

    SciTech Connect

    Reifman, J.; Vitela, J.E.; Gibbs, K.S.; Benedict, R.W.

    1995-06-01

    Two classification techniques, standard control charts and artificial neural networks, are studied as a means for automating the visual inspection of the welding of end plugs onto the top of remotely manufactured reprocessed nuclear fuel element jackets. Classificatory data are obtained through measurements performed on pre- and post-weld images captured with a remote camera and processed by an off-the-shelf vision system. The two classification methods are applied in the classification of 167 dummy stainless steel (HT9) fuel jackets yielding comparable results.

  4. Thermionic Fuel Element Verification Program - Overview

    NASA Astrophysics Data System (ADS)

    Bohl, Richard J.; Dahlberg, Richard C.; Dutt, Dale S.; Wood, John T.

    The Thermionic Fuel Element (TFE) Verification program was established in 1986 to resolve the technology concerns raised in Phase 1 of the SP-100 program, namely, the performance and lifetime of thermionic fuel elements in a fast spectrum reactor. The program builds directly on an extensive database developed in the 1960s and early 1970s in an AEC/NASA-sponsored program, when TFEs were developed and tested at design conditions for over 10,000 h. The current effort has reestablished that technology and is extending the lifetime up to 7 to 10 yr. A TFE lifetime of more than 2 yr has been demonstrated in the TRIGA reactor. Component lifetimes of more than 10 yr have been demonstrated in accelerated tests in the FFTF (Richland) and EBR-II (Idaho) test reactors. Program completion is scheduled for FY-95.

  5. 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.

  6. 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)

  7. METHOD OF MAKING WIRE FUEL ELEMENTS

    DOEpatents

    Zambrow, J.L.

    1960-08-01

    A method is given for making a nuclear reactor fuel element in the form of a uranium-bearing wire clad with zirconium. A uranium bar is enclosed in a zirconium sheath which is coated with an oxide of magnesium, beryllium, or zirconium. The sheathed bar is then placed in a steel tube and reduced to the desired diameter by swaging at 800 to 900 deg C, after which the steel and oxide are removed.

  8. 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)

  9. METHOD OF PREPARING A CERAMIC FUEL ELEMENT

    DOEpatents

    Ross, W.T.; Bloomster, C.H.; Bardsley, R.E.

    1963-09-01

    A method is described for preparing a fuel element from -325 mesh PuO/ sub 2/ and -20 mesh UO/sub 2/, and the steps of screening --325 mesh UO/sub 2/ from the -20 mesh UO/sub 2/, mixing PuO/sub 2/ with the --325 mesh UO/sub 2/, blending this mixture with sufficient --20 mesh UO/sub 2/ to obtain the desired composition, introducing the blend into a metal tube, repeating the procedure until the tube is full, and vibrating the tube to compact the powder are included. (AEC)

  10. Thermionic fuel element Verification Program - Overview

    NASA Astrophysics Data System (ADS)

    Bohl, Richard J.; Dahlberg, Richard C.; Dutt, Dale S.; Wood, John T.

    The TFE Verification Program is in the sixth year of a program to demonstrate the performance and lifetime of thermionic fuel elements for high power space applications. Data from accelerated tests in FETF and EBR-II show component lifetimes longer than 7 yr. Alumina insulators have shown good performance at high fast fluence. Graphite-cesium reservoirs based on isotropic graphite also meet requirements. Three TFEs are currently operating in the TRIGA reactor, the oldest having accumulated 15,000 hr of irradiation as of 1 October 1990.

  11. Thermionic fuel element verification program—overview

    NASA Astrophysics Data System (ADS)

    Bohl, Richard J.; Dutt, Dale S.; Dahlberg, Richard C.; Wood, John T.

    1991-01-01

    TFE Verification Program is in the sixth year of a program to demonstrate the performance and lifetime of thermionic fuel elements for high power space applications. It is jointly funded by SIDO and DOE. Data from accelerated tests in FFTF and EBR-II show component lifetimes longer than 7 years. Alumina insulators have shown good performance at high fast fluence. Graphite-cesium reservoirs based on isotropic graphite also meet requirements. Three TFEs are current operating in the TRIGA reactor, the oldest having accumulated 15,000 hours of irradiation as of 1 October 1990.

  12. Long-Term Effects of Segmental Lumbar Spinal Fusion on Adjacent Healthy Discs: A Finite Element Study

    PubMed Central

    Srinivas, Gunti Ranga; Deb, Anindya; Kurnool, Goutham

    2016-01-01

    Study Design Experimental study. Purpose The aim of the study was to develop a finite element (FE) model to study the long-term effects of various types of lumbar spinal interventions on the discs adjacent to the fused segment. Overview of Literature Earlier FE studies have been limited to one particular type of fusion and comparative quantification of the adjacent disc stresses for different types of surgical interventions has not been reported. Methods A computer aided engineering (CAE) based approach using implicit FE analysis assessed the stresses in the lumbar discs adjacent to the fused segment following anterior and posterior lumbar spine fusions at one, two and three levels (with and without instrumentation). Results It was found that instrumentation and length of fusion were the most significant factors in increasing adjacent level stresses in the lumbar discs. Conclusions In the present study, a calibrated FE model that examined spinal interventions under similar loading and boundary conditions was used to provide quantitative data which would be useful for clinicians to understand the probable long-term effect of their choice of surgical intervention. PMID:27114758

  13. Preparation of high temperature gas-cooled reactor fuel element

    DOEpatents

    Bradley, Ronnie A.; Sease, John D.

    1976-01-01

    This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

  14. 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.

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

    NASA Technical Reports Server (NTRS)

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

    1969-01-01

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

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

    DOEpatents

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

    1973-12-11

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

  17. Rapid characterization of CRISPR-Cas9 protospacer adjacent motif sequence elements.

    PubMed

    Karvelis, Tautvydas; Gasiunas, Giedrius; Young, Joshua; Bigelyte, Greta; Silanskas, Arunas; Cigan, Mark; Siksnys, Virginijus

    2015-01-01

    To expand the repertoire of Cas9s available for genome targeting, we present a new in vitro method for the simultaneous examination of guide RNA and protospacer adjacent motif (PAM) requirements. The method relies on the in vitro cleavage of plasmid libraries containing a randomized PAM as a function of Cas9-guide RNA complex concentration. Using this method, we accurately reproduce the canonical PAM preferences for Streptococcus pyogenes, Streptococcus thermophilus CRISPR3 (Sth3), and CRISPR1 (Sth1). Additionally, PAM and sgRNA solutions for a novel Cas9 protein from Brevibacillus laterosporus are provided by the assay and are demonstrated to support functional activity in vitro and in plants. PMID:26585795

  18. Fuel Element Transfer Cask Modelling Using MCNP Technique

    NASA Astrophysics Data System (ADS)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  19. Fuel Element Transfer Cask Modelling Using MCNP Technique

    SciTech Connect

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-05

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  20. 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

  1. NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

    DOEpatents

    Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

    1957-11-12

    This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

  2. Stress induced gene expression drives transient DNA methylation changes at adjacent repetitive elements

    PubMed Central

    Secco, David; Wang, Chuang; Shou, Huixia; Schultz, Matthew D; Chiarenza, Serge; Nussaume, Laurent; Ecker, Joseph R; Whelan, James; Lister, Ryan

    2015-01-01

    Cytosine DNA methylation (mC) is a genome modification that can regulate the expression of coding and non-coding genetic elements. However, little is known about the involvement of mC in response to environmental cues. Using whole genome bisulfite sequencing to assess the spatio-temporal dynamics of mC in rice grown under phosphate starvation and recovery conditions, we identified widespread phosphate starvation-induced changes in mC, preferentially localized in transposable elements (TEs) close to highly induced genes. These changes in mC occurred after changes in nearby gene transcription, were mostly DCL3a-independent, and could partially be propagated through mitosis, however no evidence of meiotic transmission was observed. Similar analyses performed in Arabidopsis revealed a very limited effect of phosphate starvation on mC, suggesting a species-specific mechanism. Overall, this suggests that TEs in proximity to environmentally induced genes are silenced via hypermethylation, and establishes the temporal hierarchy of transcriptional and epigenomic changes in response to stress. DOI: http://dx.doi.org/10.7554/eLife.09343.001 PMID:26196146

  3. Fuel cell elements with improved water handling capacity

    NASA Technical Reports Server (NTRS)

    Kindler, Andrew (Inventor); Lee, Albany (Inventor)

    2001-01-01

    New fuel cell components for use in liquid feed fuel cell systems are provided. The components include biplates and endplates, having a hydrophilic surface and allow high efficiency operation. Conductive elements and a wicking device also form a part of the fuel cell components of the invention.

  4. 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.

  5. IN-CELL visual examinations of K east fuel elements

    SciTech Connect

    Pitner, A.L.; Pyecha, T.D., Fluor Daniel Hanford

    1997-03-06

    Nine outer fuel elements were recovered from the K East Basin and transferred to a hot cell for examination. Extensive testing planned for these elements will support the process design for the Integrated Process Strategy (IPS), with emphasis on drying and conditioning behavior. Visual examinations of the fuel elements confirmed that they are appropriate to meet testing objectives to provide design guidance for IPS processing parameters.

  6. Metal elements in the bottom sediments of the Changjiang Estuary and its adjacent continental shelf of the East China Sea.

    PubMed

    Cao, Lu; Hong, Gi Hoon; Liu, Sumei

    2015-06-15

    The metal elements (Al, Fe, Mn, Cr, Co, Ni, Cu, Zn, As, Pb and Ca) in the bottom sediment of the Changjiang Estuary and its adjacent continental shelf of the East China Sea were studied to map their spatial distribution and to assess their potential risk to the marine biota. These metal concentrations except Ca were generally higher in the inner shelf and northeastern part, and were found to decrease from the coast to the offshore of the Changjiang Estuary. Sedimentary Ca was most abundant in the outer shelf sediments and decreased in inner shelf. Arsenic (As) appeared to be contaminated due to economic development from 1980s in the inner shelf overall, but the potential ecological risk from the selected metals was low in the coastal sea off the Changjiang. PMID:25869200

  7. VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

    DOEpatents

    Furgerson, W.T.

    1963-12-17

    A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

  8. Nuclear fuel elements and method of making same

    DOEpatents

    Schweitzer, Donald G.

    1992-01-01

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

  9. NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

    DOEpatents

    Brooks, H.

    1960-04-26

    A description is given for a fuel element comprising a body of uranium metal or an uranium compound dispersed in a matrix material made from magnesium, calcium, or barium and a stainless steel jacket enclosing the body.

  10. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    SciTech Connect

    N.E. Woolstenhulme; M.K. Meyer; D.M. Wachs

    2011-10-01

    The US fuel development team has performed numerous irradiation tests on small to medium sized specimens containing low enriched uranium fuel designs. The team is now focused on qualification and demonstration of the uranium-molybdenum Base Monolithic Design and has entered the next generation of testing with the design and irradiation of prototypic elements which contain this fuel. The designs of fuel elements containing monolithic fuel, such as AFIP-7 (which is currently under irradiation) and RERTR-FE (which is currently under fabrication), are appropriate progressions relative to the technology life cycle. The culmination of this testing program will occur with the design, fabrication, and irradiation of demonstration products to include the base fuel demonstration and design demonstration experiments. Future plans show that design, fabrication, and testing activities will apply the rigor needed for a demonstration campaign.

  11. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOEpatents

    Flint, O.

    1961-01-10

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

  12. Design and experimental investigation into fuel element melting during pulsed heating in the IGRIK

    SciTech Connect

    Levakov, B.G.; Andreev, V.V.; Vasilyev, A.P.

    1995-12-31

    Research has been performed on reactor fuel melting with pulsed input of energy in fuel elements up to 1.3 kj/g. The following were determined: energy input in fuel elements and energy input tempo; fission number distribution by the radius of the fuel element; the temperature of fuel and ampoule walls; and displacement of fuel boundaries.

  13. The quantification of mixture stoichiometry when fuel molecules contain oxidizer elements or oxidizer molecules contain fuel elements.

    SciTech Connect

    Mueller, Charles J.

    2005-05-01

    The accurate quantification and control of mixture stoichiometry is critical in many applications using new combustion strategies and fuels (e.g., homogeneous charge compression ignition, gasoline direct injection, and oxygenated fuels). The parameter typically used to quantify mixture stoichiometry (i.e., the proximity of a reactant mixture to its stoichiometric condition) is the equivalence ratio, /gf. The traditional definition of /gf is based on the relative amounts of fuel and oxidizer molecules in a mixture. This definition provides an accurate measure of mixture stoichiometry when the fuel molecule does not contain oxidizer elements and when the oxidizer molecule does not contain fuel elements. However, the traditional definition of /gf leads to problems when the fuel molecule contains an oxidizer element, as is the case when an oxygenated fuel is used, or once reactions have started and the fuel has begun to oxidize. The problems arise because an oxidizer element in a fuel molecule is counted as part of the fuel, even though it acts as an oxidizer. Similarly, if an oxidizer molecule contains fuel elements, the fuel elements in the oxidizer molecule are misleadingly lumped in with the oxidizer in the traditional definition of /gf. In either case, use of the traditional definition of /gf to quantify the mixture stoichiometry can lead to significant errors. This paper introduces the oxygen equivalence ratio, /gf/gV, a parameter that properly characterizes the instantaneous mixture stoichiometry for a broader class of reactant mixtures than does /gf. Because it is an instantaneous measure of mixture stoichiometry,/gf/gV can be used to track the time-evolution of stoichiometry as a reaction progresses. The relationship between /gf/gV and /gf is shown. Errors are involved when the traditional definition of /gf is used as a measure of mixture stoichiometry with fuels that contain oxidizer elements or oxidizers that contain fuel elements; /gf/gV is used to quantify

  14. Nuclear breeder reactor fuel element with silicon carbide getter

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1987-01-01

    An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

  15. PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

    DOEpatents

    Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

    1963-09-01

    A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

  16. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

    1993-01-14

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

  17. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y. )

    1993-01-15

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of [similar to]60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of [similar to]100 MW/L may be achievable.

  18. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

    1993-05-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

  19. Trace Element Mobility in Water and Sediments in a Hyporheic Zone Adjacent to an Abandoned Uranium Mine

    NASA Astrophysics Data System (ADS)

    Roldan, C.; Blake, J.; Cerrato, J.; Ali, A.; Cabaniss, S.

    2015-12-01

    The legacy of abandoned uranium mines lead to community concerns about environmental and health effects. This study focuses on a cross section of the Rio Paguate, adjacent to the Jackpile Mine on the Laguna Reservation, west-central New Mexico. Often, the geochemical interactions that occur in the hyporheic zone adjacent to these abandoned mines play an important role in trace element mobility. In order to understand the mobility of uranium (U), arsenic (As), and vanadium (V) in the Rio Paguate; surface water, hyporheic zone water, and core sediment samples were analyzed using inductively coupled plasma mass spectroscopy (ICP-MS). All water samples were filtered through 0.45μm and 0.22μm filters and analyzed. The results show that there is no major difference in concentrations of U (378-496μg/L), As (0.872-6.78μg/L), and V (2.94-5.01μg/L) between the filter sizes or with depth (8cm and 15cm) in the hyporheic zone. The unfiltered hyporheic zone water samples were analyzed after acid digestion to assess the particulate fraction. These results show a decrease in U concentration (153-202μg/L) and an increase in As (33.2-219μg/L) and V (169-1130μg/L) concentrations compared to the filtered waters. Surface water concentrations of U(171-184μg/L) are lower than the filtered hyporheic zone waters while As(1.32-8.68μg/L) and V(1.75-2.38μg/L) are significantly lower than the hyporheic zone waters and particulates combined. Concentrations of As in the sediment core samples are higher in the first 15cm below the water-sediment interface (14.3-3.82μg/L) and decrease (0.382μg/L) with depth. Uranium concentrations are consistent (0.047-0.050μg/L) at all depths. The over all data suggest that U is mobile in the dissolved phase and both As and V are mobile in the particular phase as they travel through the system.

  20. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    NASA Astrophysics Data System (ADS)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  1. Local Burn-Up Effects in the NBSR Fuel Element

    SciTech Connect

    Brown N. R.; Hanson A.; Diamond, D.

    2013-01-31

    This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peaking relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.

  2. Analysis of the ATR fuel element swaging process

    SciTech Connect

    Richins, W.D.; Miller, G.K.

    1995-12-01

    This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B&W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF.

  3. The manufacture of LEU fuel elements at Dounreay

    SciTech Connect

    Gibson, J.

    1997-08-01

    Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

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

    NASA Technical Reports Server (NTRS)

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

    2011-01-01

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

  5. Drying damaged K West fuel elements (Summary of whole element furnace runs 1 through 8)

    SciTech Connect

    LAWRENCE, L.A.

    1998-10-13

    N Reactor fuel elements stored in the Hanford K Basins were subjected to high temperatures and vacuum conditions to remove water. Results of the first series of whole element furnace tests i.e., Runs 1 through 8 were collected in this summary report. The report focuses on the six tests with breached fuel from the K West Basin which ranged from a simple fracture at the approximate mid-point to severe damage with cladding breaches at the top and bottom ends with axial breaches and fuel loss. Results of the tests are summarized and compared for moisture released during cold vacuum drying, moisture remaining after drying, effects of drying on the fuel element condition, and hydrogen and fission product release.

  6. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Loeb, E.; Nicklas, J.H.

    1959-02-01

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

  7. Cryogenic Thermal Expansion of Y-12 Graphite Fuel Elements

    SciTech Connect

    Eash, D. T.

    2013-07-08

    Thermal expansion measurements betwccn 20°K and 300°K were made on segments of three uranium-loaded Y-12 uncoated graphite fuel elements. The thermal expansion of these fuel elements over this temperature range is represented by the equation: {Delta}L/L = -39.42 x 10{sup -5} + 1.10 x 10{sup -7} T + 6.47 x 10{sup -9} T{sup 2} - 8.30 x 10{sup -12} T{sup 3}.

  8. Method and apparatus for diagnosing breached fuel elements

    DOEpatents

    Gross, K.C.; Lambert, J.D.B.; Nomura, S.

    1987-03-02

    The invention provides an apparatus and method for diagnosing breached fuel elements in a nuclear reactor. A detection system measures the activity of isotopes from the cover gas in the reactor. A data acquisition and processing system monitors the detection system and corrects for the effects of the cover-gas clean up system on the measured activity and further calculates the derivative curve of the corrected activity as a function of time. A plotting system graphs the derivative curve, which represents the instantaneous release rate of fission gas from a breached fuel element. 8 figs.

  9. Expert system for surveillance and diagnosis of breach fuel elements

    DOEpatents

    Gross, K.C.

    1988-01-21

    An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil area of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor. 2 figs.

  10. Expert system for surveillance and diagnosis of breach fuel elements

    DOEpatents

    Gross, Kenny C.

    1989-01-01

    An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor.

  11. Method and apparatus for diagnosing breached fuel elements

    DOEpatents

    Gross, Kenny C.; Lambert, John D. B.; Nomura, Shigeo

    1988-01-01

    The invention provides an apparatus and method for diagnosing breached fuel elements in a nuclear reactor. A detection system measures the activity of isotopes from the cover-gas in the reactor. A data acquisition and processing system monitors the detection system and corrects for the effects of the cover-gas clean up system on the measured activity and further calculates the derivative cure of the corrected activity as a function of time. A plotting system graphs the derivative curve, which represents the instantaneous release rate of fission gas from a breached fuel element.

  12. 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.

  13. NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

    DOEpatents

    Kingston, W.E.; Kopelman, B.; Hausner, H.H.

    1963-07-01

    A fuel element consisting of uranium nitride and uranium carbide in the form of discrete particles in a solid coherent matrix of a metal such as steel, beryllium, uranium, or zirconium and clad with a metal such as steel, aluminum, zirconium, or beryllium is described. The element is made by mixing powdered uranium nitride and uranium carbide with powdered matrix metal, then compacting and sintering the mixture. (AEC)

  14. Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

    NASA Technical Reports Server (NTRS)

    Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

    2013-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

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

    DOEpatents

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

    1962-12-01

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

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

    DOEpatents

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

    1962-01-23

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

  17. 34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT FRAME. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-4. INEL INDEX CODE NUMBER: 075 0701 60 851 151978. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

  18. 36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT FRAME AND SUPPORT PLATFORM, AND SAFETY MECHANISM ASSEMBLY (SPRING-LOADED HINGE). F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-1. INEL INDEX CODE NUMBER: 075 0701 60 851 151975. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

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

    NASA Technical Reports Server (NTRS)

    Moran, Robert P.; Emrich, William

    2011-01-01

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

  20. Some parametric flow analyses of a particle bed fuel element

    SciTech Connect

    Dobranich, D.

    1993-05-01

    Parametric calculations are performed, using the SAFSIM computer program, to investigate the fluid mechanics and heat transfer performance of a particle bed fuel element. Both steady-state and transient calculations are included, addressing such issues as flow stability, reduced thrust operation, transpiration drag, coolant conductivity enhancement, flow maldistributions, decay heat removal, flow perturbations, and pulse cooling. The calculations demonstrate the dependence of the predicted results on the modeling assumptions and thus provide guidance as to where further experimental and computational investigations are needed. The calculations also demonstrate that both flow instability and flow maldistribution in the fuel element are important phenomena. Furthermore, results are encouraging that geometric design changes to the element can significantly reduce problems related to these phenomena, allowing improved performance over a wide range of element power densities and flow rates. Such design changes will help to maximize the operational efficiency of space propulsion reactors employing particle bed fuel element technology. Finally, the results demonstrate that SAFSIM is a valuable engineering tool for performing quick and inexpensive parametric simulations addressing complex flow problems.

  1. The OSU Hydro-Mechanical Fuel Test Facility: Standard Fuel Element Testing

    SciTech Connect

    Wade R. Marcum; Brian G. Woods; Ann Marie Phillips; Richard G. Ambrosek; James D. Wiest; Daniel M. Wachs

    2001-10-01

    Oregon State University (OSU) and the Idaho National Laboratory (INL) are currently collaborating on a test program which entails hydro-mechanical testing of a generic plate type fuel element, or standard fuel element (SFE), for the purpose of qualitatively demonstrating mechanical integrity of uranium-molybdenum monolithic plates as compared to that of uranium aluminum dispersion, and aluminum fuel plates due to hydraulic forces. This test program supports ongoing work conducted for/by the fuel development program and will take place at OSU in the Hydro-Mechanical Fuel Test Facility (HMFTF). Discussion of a preliminary test matrix, SFE design, measurement and instrumentation techniques, and facility description are detailed in this paper.

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

    NASA Technical Reports Server (NTRS)

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

    2011-01-01

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

  3. Method of locating a leaking fuel element in a fast breeder power reactor

    DOEpatents

    Honekamp, John R.; Fryer, Richard M.

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  4. FISSILE MATERIAL AND FUEL ELEMENTS FOR NEUTRONIC REACTORS

    DOEpatents

    Shaner, B.E.

    1961-08-15

    The fissile material consists of about 64 to 70% (weight) zirconium dioxide, 15 to 19% uranium dioxide, and 8 to 17% calcium oxide. The fissile material is formed into sintered composites which are disposed in a compartmented fuel element, comprising essentially a flat filler plate having a plurality of compartments therein, enclosed in cladding plates of the same material as the filler plate. The resultant fuel has good resistance to corrosion in high temperature pressurized water, good dimensional stability to elevated temperatures, and good resistance to thermal shock. (AEC)

  5. Nuclear reactor fuel element with vanadium getter on cladding

    DOEpatents

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

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

  6. Method for measuring recovery of catalytic elements from fuel cells

    DOEpatents

    Shore, Lawrence; Matlin, Ramail

    2011-03-08

    A method is provided for measuring the concentration of a catalytic clement in a fuel cell powder. The method includes depositing on a porous substrate at least one layer of a powder mixture comprising the fuel cell powder and an internal standard material, ablating a sample of the powder mixture using a laser, and vaporizing the sample using an inductively coupled plasma. A normalized concentration of catalytic element in the sample is determined by quantifying the intensity of a first signal correlated to the amount of catalytic element in the sample, quantifying the intensity of a second signal correlated to the amount of internal standard material in the sample, and using a ratio of the first signal intensity to the second signal intensity to cancel out the effects of sample size.

  7. METHOD AND APPARATUS FOR EXAMINING FUEL ELEMENTS FOR LEAKAGE

    DOEpatents

    Smith, R.R.; Echo, M.W.; Doe, C.B.

    1963-12-31

    A process and a device for the continuous monitoring of fuel elements while in use in a liquid-metal-cooled, argonblanketed nuclear reactor are presented. A fraction of the argon gas is withdrawn, contacted with a negative electrical charge for attraction of any alkali metal formed from argon by neutron reaction, and recycled into the reactor. The electrical charge is introduced into water, and the water is examined for radioactive alkali metals. (AEC)

  8. FUEL ELEMENTS FOR NUCLEAR REACTORS AND PROCESS OF MAKING

    DOEpatents

    Roake, W.E.

    1958-08-19

    A process is described for producing uranium metal granules for use in reactor fuel elements. The granules are made by suspending powdered uramiunn metal or uranium hydride in a viscous, non-reactive liquid, such as paraffin oil, aad pouring the resulting suspension in droplet, on to a bed of powdered absorbent. In this manner the liquid vehicle is taken up by the sorbent and spherical pellets of uranium metal are obtained. The

  9. Selection of Isotopes and Elements for Fuel Cycle Analysis

    SciTech Connect

    Steven J. Piet

    2009-04-01

    Fuel cycle system analysis simulations examine how the selection among fuel cycle options for reactors, fuel, separation, and waste management impact uranium ore utilization, waste masses and volumes, radiotoxicity, heat to geologic repositories, isotope-dependent proliferation resistance measures, and so forth. Previously, such simulations have tended to track only a few actinide and fission product isotopes, those that have been identified as important to a few criteria from the standpoint of recycled material or waste, taken as a whole. After accounting for such isotopes, the residual mass is often characterized as “fission product other” or “actinide other”. However, detailed assessment of separation and waste management options now require identification of key isotopes and residual mass for Group 1A/2A elements (Rb, Cs, Sr, Ba), inert gases (Kr, Xe), halogens (Br, I), lanthanides, transition metals, transuranic (TRU), uranium, actinide decay products. The paper explains the rationale for a list of 81 isotopes and chemical elements to better support separation and waste management assessment in dynamic system analysis models such as Verifiable Fuel Cycle Simulation (VISION)

  10. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    SciTech Connect

    Knight, R.W.; Morin, R.A.

    1999-12-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

  11. Coupled thermionic and thermalhydraulic analyses of thermionic fuel elements

    NASA Astrophysics Data System (ADS)

    Pawlowski, Ronald A.; Klein, Andrew C.; McVey, John B.

    The authors discuss a heat transfer analysis of a 'single cell' TFE (thermionic fuel element), that is, within the TFE a single emitter and collector cover the entire length of the UO2 fuel (approximately 25 cm). The electrical conversion performance of the TFE is investigated for a range of operating conditions. The dependence of maximum fuel temperature on the TFE operating parameters, such as total thermal power, current output, and coolant inlet temperature, is also discussed. A computer code (TFEHX) to model the thermal and electrical performance of the TFE has been developed. The results from the TFEHX code consist of a wide range of TFE operational parameters, including the temperature distributions within the TFE, the overall electrical power output, the conversion efficiency, the voltage difference between the electrode leads, the electrical losses and the ohmic heating in the electrodes, and the coolant temperature profile. Results from this code indicate that a single-celled TFE is more efficient and is less likely to experience melting of its fuel if a uniform amount of heat is generated along its length.

  12. Decisive factor in increase of loading at adjacent segments after lumbar fusion: operative technique, pedicle screws, or fusion itself: biomechanical analysis using finite element

    NASA Astrophysics Data System (ADS)

    Park, Joon-Hee; Kim, Ho-Joong; Kang, Kyoung-Tak; Kim, Ka-yeon; Chun, Heoung-Jae; Moon, Seong-Hwan; Lee, Hwan-Mo

    2010-03-01

    The aim of this study is to investigate the change in biomechanical milieu following removal of pedicle screws or removal of spinous process with posterior ligament complex in instrumented single level lumbar arthrodesis. We developed and validated a finite element model (FEM) of the intact lumbar spine (L2-4). Four scenarios of L3-4 lumbar fusion were simulated: posterolateral fusion (PLF) at L3-4 using pedicle screw system with preservation of PLC (Pp WiP), L3-4 lumbar posterolateral fusion state after removal of pedicle screw system with preservation of PLC (Pp WoP), L3-4 using pedicle screw system without preservation PLC (Sp WiP), L3-4 lumbar posterolateral fusion state after removal of pedicle screw system without preservation of PLC (Sp WoP). For these models, we investigated the range of motion and maximal Von mises stress of disc in all segments under various moments. All fusion models demonstrated increase in range of motion at adjacent segments compared to the intact model.For the four fusion models, the WiP model s P had the largest increase in range of motion at each adjacent segment. This study demonstrated that removal of pedicle screw system and preservation of PLC after complete lumbar spinal fusion could reduce the stress of adjacent segments synergistically and might have beneficial effects in preventing ASD.

  13. Decisive factor in increase of loading at adjacent segments after lumbar fusion: operative technique, pedicle screws, or fusion itself: biomechanical analysis using finite element

    NASA Astrophysics Data System (ADS)

    Park, Joon-Hee; Kim, Ho-Joong; Kang, Kyoung-Tak; Kim, Ka-Yeon; Chun, Heoung-Jae; Moon, Seong-Hwan; Lee, Hwan-Mo

    2009-12-01

    The aim of this study is to investigate the change in biomechanical milieu following removal of pedicle screws or removal of spinous process with posterior ligament complex in instrumented single level lumbar arthrodesis. We developed and validated a finite element model (FEM) of the intact lumbar spine (L2-4). Four scenarios of L3-4 lumbar fusion were simulated: posterolateral fusion (PLF) at L3-4 using pedicle screw system with preservation of PLC (Pp WiP), L3-4 lumbar posterolateral fusion state after removal of pedicle screw system with preservation of PLC (Pp WoP), L3-4 using pedicle screw system without preservation PLC (Sp WiP), L3-4 lumbar posterolateral fusion state after removal of pedicle screw system without preservation of PLC (Sp WoP). For these models, we investigated the range of motion and maximal Von mises stress of disc in all segments under various moments. All fusion models demonstrated increase in range of motion at adjacent segments compared to the intact model.For the four fusion models, the WiP model s P had the largest increase in range of motion at each adjacent segment. This study demonstrated that removal of pedicle screw system and preservation of PLC after complete lumbar spinal fusion could reduce the stress of adjacent segments synergistically and might have beneficial effects in preventing ASD.

  14. Elements of a general theory of joints. 7. Mechanical structures of the relative motion of adjacent vertebrae.

    PubMed

    Nägerl, H; Kubein-Meesenburg, D; Fanghänel, J

    1992-02-01

    The mechanical structure of the relative motion of adjacent vertebrae is given by the interplay of the artt. zygapophysiales, the intervertebral disk and the ligaments, respectively. This structure will be described in the 3 main functional planes. Horizontal plane (parallel to the disk): The two artt. zygapophysiales structurally represent a force locked, kinematically contrained four-bar-chain (link quadrangle). It possesses but one degree of freedom. The intervertebral disk does not influence, therefore, the kinematic structure but the extent of the motion. Frontal-coronal plane: The four-bar-chain of the frontal plane seems to be inflexible. Lateral bending is only possible combined with a right or left handed axial rotation. Vertical-sagittal plane: The artt. zygapophysiales work as a single dimeric link chain. The momentary rotational center lies on the common normal of the two articular surfaces. Its position upon that is determined by the geometry of the applied force system according to the "reciprocal theorem of synarthroses". Solely one direction of translation is possible. A translation perpendicular to the intervertebral disk is impossible in the cervical and thoracic region. PMID:1605362

  15. Triaxial Swirl Injector Element for Liquid-Fueled Engines

    NASA Technical Reports Server (NTRS)

    Muss, Jeff

    2010-01-01

    A triaxial injector is a single bi-propellant injection element located at the center of the injector body. The injector element consists of three nested, hydraulic swirl injectors. A small portion of the total fuel is injected through the central hydraulic injector, all of the oxidizer is injected through the middle concentric hydraulic swirl injector, and the balance of the fuel is injected through an outer concentric injection system. The configuration has been shown to provide good flame stabilization and the desired fuel-rich wall boundary condition. The injector design is well suited for preburner applications. Preburner injectors operate at extreme oxygen-to-fuel mass ratios, either very rich or very lean. The goal of a preburner is to create a uniform drive gas for the turbomachinery, while carefully controlling the temperature so as not to stress or damage turbine blades. The triaxial injector concept permits the lean propellant to be sandwiched between two layers of the rich propellant, while the hydraulic atomization characteristics of the swirl injectors promote interpropellant mixing and, ultimately, good combustion efficiency. This innovation is suited to a wide range of liquid oxidizer and liquid fuels, including hydrogen, methane, and kerosene. Prototype testing with the triaxial swirl injector demonstrated excellent injector and combustion chamber thermal compatibility and good combustion performance, both at levels far superior to a pintle injector. Initial testing with the prototype injector demonstrated over 96-percent combustion efficiency. The design showed excellent high -frequency combustion stability characteristics with oxygen and kerosene propellants. Unlike the more conventional pintle injector, there is not a large bluff body that must be cooled. The absence of a protruding center body enhances the thermal durability of the triaxial swirl injector. The hydraulic atomization characteristics of the innovation allow the design to be

  16. Water-quality, water-level, and lake-bottom-sediment data collected from the defense fuel supply point and adjacent properties, Hanahan, South Carolina, 1990-96

    USGS Publications Warehouse

    Petkewich, M.D.; Vroblesky, D.A.; Robertson, J.F.; Bradley, P.M.

    1997-01-01

    A 9-year scientific investigation to determine the potential for biore-mediation of ground-water contamination and to monitor the effectiveness of an engineered bioremediation system located at the Defense Fuel Supply Point and adjacent properties in Hanahan, S.C., has culminated in the collection of abundant water-quality and water-level data.This report presents the analytical results of the study that monitored the changes in surface- and ground-water quality and water-table elevations in the study area from December 1990 to January 1996. This report also presents analytical results of lake-bottom sediments collected in the study area.

  17. Corrosion studies in fuel element reprocessing environments containing nitric acid

    SciTech Connect

    Beavers, J A; White, R R; Berry, W E; Griess, J C

    1982-04-01

    Nitric acid is universally used in aqueous fuel element reprocessing plants; however, in the processing scheme being developed by the Consolidated Fuel Reprocessing Program, some of the equipment will be exposed to nitric acid under conditions not previously encountered in fuel element reprocessing plants. A previous report presented corrosion data obtained in hyperazeotropic nitric acid and in concentrated magnesium nitrate solutions used in its preparation. The results presented in this report are concerned with the following: (1) corrosion of titanium in nitric acid; (2) corrosion of nickel-base alloys in a nitric acid-hydrofluoric acid solution; (3) the formation of Cr(VI), which enhances corrosion, in nitric acid solutions; and (4) corrosion of mechanical pipe connectors in nitric acid. The results show that the corrosion rate of titanium increased with the refreshment rate of boiling nitric acid, but the effect diminished rapidly as the temperature decreased. The addition of iodic acid inhibited attack. Also, up to 200 ppM of fluoride in 70% HNO/sub 3/ had no major effect on the corrosion of either titanium or tantalum. In boiling 8 M HNO/sub 3/-0.05 M HF, Inconel 671 was more resistant than Inconel 690, but both alloys experienced end-grain attack. In the case of Inconel 671, heat treatment was very important; annealed and quenched material was much more resistant than furnace-cooled material.The rate of oxidation of Cr(III) to Cr(VI) increased significantly as the nitric acid concentration increased, and certain forms of ruthenium in the solution seemed to accelerate the rate of formation. Mechanical connectors of T-304L stainless steel experienced end-grain attack on the exposed pipe ends, and seal rings of both stainless steel and a titanium alloy (6% Al-4% V) underwent heavy attack in boiling 8 M HNO/sub 3/.

  18. 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

  19. From adjacent activation in Escherichia coli and DNA cyclization to eukaryotic enhancers: the elements of a puzzle

    PubMed Central

    Amouyal, Michèle

    2014-01-01

    auxiliary elements known as insulators. PMID:25404937

  20. The effect of coprecipitation in some key spent fuel elements

    NASA Astrophysics Data System (ADS)

    Quiñones, J.; Serrano, J.; Diaz Arocas, P.

    2001-09-01

    Performance assessment (PA) of high-level waste (HLW) repositories needs to know real aqueous concentrations of key radionuclides under repository conditions for assuring the safety of the emplacement. The scarcity of these values under repository conditions leads to the use, in the PA studies, of the solubility of pure phases, which is a conservative assumption. Coprecipitation experiments are a very useful tool for giving realistic solubilities of key radionuclides. In this work, experimental data obtained from spent fuel (SF) and SIMFUEL coprecipitation tests under granite and saline conditions are presented. The experimental concentrations measured for several elements when equilibrium was achieved were much lower than expected considering only the solubility of pure phases. To explain this discrepancy, a tentative approach for modelling these experimental leaching and precipitation results of uranium, plutonium, americium, and strontium taking into account solid solution formations was made.

  1. Analysis of Ya-21u thermionic fuel elements

    SciTech Connect

    Paramonov, D.V.; El-Genk, M.S.

    1996-12-01

    The Ya-21u unit of the Soviet-made TOPAZ-II power system has recently been tested at the Thermionic Evaluation Facility in Albuquerque, New Mexico. A change in the unit performance was measured during these tests. In an attempt to identify the causes of this change performance, data were examined and used to estimate surface properties of electrodes of thermionic fuel elements (TFEs) of the power system. The effective emissivity was estimated at {approximately}0.03 to 0.035 higher than for as-fabricated TFE and cesiated work functions of the electrodes, which were higher than for as-fabricated TFEs. These changes in the effective emissivity and cesiated work functions, caused by gaseous impurities and air incursion in the TFEs interelectrode gap, lowered both the emitter temperature and the output load voltage thus contributing to the measured decrease in output power.

  2. Thermionic Fuel Element performance: TFE Verification Program. Final test report

    SciTech Connect

    Not Available

    1994-06-01

    The program objective is to demonstrate the technology readiness of a Thermionic Fuel Element (TFE) suitable for use as the basic element in a thermionic reactor with electric power output in the 0.5 to 5.0 MW(e) range, and a full power life of 7 years. A TFE was designed that met the reliability and lifetime requirements for a 2 MW(e) conceptual reactor design. Analysis showed that this TFE could be used over the range of 0.5 to 5 megawatts. This was used as the basis for designing components for test and evaluation. The demonstration of a 7-year component lifetime capability was through the combined use of analytical models and accelerated, confirmatory tests in a fast test reactor. Iterative testing was performed in which the results of one test series led to evolutionary improvements in the next test specimens. The TFE components underwent screening and initial development testing in ex-reactor tests. Several design and materials options were considered for each component. As screening tests permitted, down selection occurred to very specific designs and materials. In parallel with ex-reactor testing, and fast reactor component testing, components were integrated into a TFE and tested in the TRIGA test reactor at GA. Realtime testing of partial length TFEs was used to test support, alignment and interconnective TFE components, and to verify TFE performance in-reactor with integral cesium reservoirs. Realtime testing was also used to verify the relation between TFE performance and fueled emitter swelling, to test the durability of intercell insulation, to check temperature distributions, and to verify the adequacy over time of the fission gas venting channels. Predictions of TFE lifetime rested primarily on the accelerated component testing results, as correlated and extended to realtime by the use of analytical models.

  3. Thermal breeder fuel enrichment zoning

    DOEpatents

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  4. 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.

  5. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    SciTech Connect

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-02-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses.

  6. Description of fuel element brush assembly`s fabrication for 105-K west

    SciTech Connect

    Maassen, D. P.

    1997-10-15

    This report is a description of the process to redesign and fabricate, as well as, describe the features of the Fuel Element Brush Assembly used in the 105-K West Basin. This narrative description will identify problems that occurred during the redesigning and fabrication of the 105-K West Basin Fuel Element Brush Assembly and specifically address their solutions.

  7. Ground test facilities for evaluating nuclear thermal propulsion engines and fuel elements

    SciTech Connect

    Allen, G.C.; Beck, D.F.; Harmon, C.D.; Shipers, L.R.

    1992-08-01

    Interagency panels evaluating nuclear thermal propulsion development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and design issues of a proposed ground test complex for evaluating nuclear thermal propulsion engines and fuel elements being developed for the Space Nuclear Thermal Propulsion (SNTP) program. 2 refs.

  8. Nuclear breeder reactor fuel element with axial tandem stacking and getter

    DOEpatents

    Gibby, Ronald L.; Lawrence, Leo A.; Woodley, Robert E.; Wilson, Charles N.; Weber, Edward T.; Johnson, Carl E.

    1981-01-01

    A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

  9. Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

    NASA Technical Reports Server (NTRS)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2015-01-01

    This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

  10. Conversion and evaluation of the THOR reactor core to TRIGA fuel elements

    SciTech Connect

    Li, S.-H.; Shiau, L.-C.

    1990-07-01

    The THOR reactor is a pool type 1 MW research reactor and has been operated since 1961. The original MTR fuel elements have been gradually replaced by TRIGA fuel elements since 1977 and the conversion completed in 1987. The calculations were performed for various core configurations by using computer codes, WIMS/CITATION. The computing results have been evaluated and compared with the core measurements after the fuel conversion. The analysis results are in good correspondence with the measurements. (author)

  11. Drying Results of K-Basin Fuel Element 2660M (Run 7)

    SciTech Connect

    B.M. Oliver; G.S. Klinger; J. Abrefah; S.C. Marschman; P.J. MacFarlan; G.A. Ritter

    1999-07-26

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the seventh of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 2660M. This element (referred to as Element 2660M) was stored underwater in the K-West Basin from 1983 until 1996. Element 2660M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0.

  12. ZPPR FUEL ELEMENT THERMAL STRESS-STRAIN ANALYSIS

    SciTech Connect

    Charles W. Solbrig; Jason Andrus; Chad Pope

    2014-04-01

    The design temperature of high plutonium concentration ZPPR fuel assemblies is 600 degrees C. Cladding integrity of the 304L stainless steel cladding is a significant concern with this fuel since even small holes can lead to substantial fuel degradation. Since the fuel has a higher coefficient of thermal expansion than the cladding, an investigation of the stress induced in the cladding due to the differential thermal expansion of fuel and cladding up to the design temperature was conducted. Small holes in the cladding envelope would be expected to lead to the fuel hydriding and oxidizing into a powder over a long period of time. This is the same type of chemical reaction chain that exists in the degradion of the high uranium concentration ZPPR fuel. Unfortunately, the uranium fuel was designed with vents which allowed this degradation to occur. The Pu cladding is sealed so only fuel with damaged cladding would be subject to this damage. The thermal stresses that can be developed in the fuel cladding have been calculated in in this paper and compared to the ultimate tensile stress of the cladding. The conclusion is drawn that thermal stresses cannot induce holes in the cladding even for the highest storage temperatures predicted in calculations (292°C). In fact, thermal stress can not cause cladding failure as long as the fuel temperatures are below the design limit of 600 degrees C (1,112 degrees F).

  13. Measurement of dynamic interaction between a vibrating fuel element and its support

    SciTech Connect

    Fisher, N.J.; Tromp, J.H.; Smith, B.A.W.

    1996-12-01

    Flow-induced vibration of CANDU{reg_sign} fuel can result in fretting damage of the fuel and its support. A WOrk-Rate Measuring Station (WORMS) was developed to measure the relative motion and contact forces between a vibrating fuel element and its support. The fixture consists of a small piece of support structure mounted on a micrometer stage. This arrangement permits position of the support relative to the fuel element to be controlled to within {+-} {micro}m. A piezoelectric triaxial load washer is positioned between the support and micrometer stage to measure contact forces, and a pair of miniature eddy-current displacement probes are mounted on the stage to measure fuel element-to-support relative motion. WORMS has been utilized to measure dynamic contact forces, relative displacements and work-rates between a vibrating fuel element and its support. For these tests, the fuel element was excited with broadband random force excitation to simulate flow-induced vibration due to axial flow. The relationship between fuel element-to-support gap or preload (i.e., interference or negative gap) and dynamic interaction (i.e., relative motion, contact forces and work-rates) was derived. These measurements confirmed numerical simulations of in-reactor interaction predicted earlier using the VIBIC code.

  14. 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.

  15. Finite element analysis of monolithic solid oxide fuel cells

    SciTech Connect

    Saigal, A. . Dept. of Mechanical Engineering); Majumdar, S. )

    1992-01-01

    This paper investigates the stress and fracture behavior of a monolithic solid oxide fuel cell (MSOFC) currently under joint development by Allied Signal Corporation and Argonne National Laboratory. The MSOFC is an all-ceramic fuel cell capable of high power density and tolerant of a variety of hydrocarbon fuels, making it potentially attractive for stationary utility and mobile transportation systems. The monolithic design eliminates inactive structural supports, increases active surface area, and lowers voltage losses caused by internal resistance.

  16. Finite element analysis of monolithic solid oxide fuel cells

    SciTech Connect

    Saigal, A.; Majumdar, S.

    1992-04-01

    This paper investigates the stress and fracture behavior of a monolithic solid oxide fuel cell (MSOFC) currently under joint development by Allied Signal Corporation and Argonne National Laboratory. The MSOFC is an all-ceramic fuel cell capable of high power density and tolerant of a variety of hydrocarbon fuels, making it potentially attractive for stationary utility and mobile transportation systems. The monolithic design eliminates inactive structural supports, increases active surface area, and lowers voltage losses caused by internal resistance.

  17. Neutron and gamma radiography of UO{sub 2} and TRIGA fuel elements

    SciTech Connect

    Robinson, A.H.; Gao, Y.C.; Johnson, A.G.; Ringle, J.C.

    1982-07-01

    The Oregon State TRIGA Reactor neutron radiography facility has been used to produce both neutron and gamma radiographs of reactor fuel. In this paper a comparison of the applicability of neutron and gamma radiography to both UO{sub 2} fuel pins and TRIGA fuel elements is made. In the case of UO{sub 2} fuel, conventional thermal neutron radiography produces excellent quality radiographs. These radiographs may be used to detect various defects in the fuel such as enrichment differences, cracks, end-capping, inclusions, etc. For TRIGA fuel elements, conventional thermal neutron radiography will not show the internal structure. This is due to the high hydrogen content of the fuel. These elements are typically 8.5 w/o uranium in Zr-H{sub 1.7}; the density of hydrogen in the fuel being about 80 percent that of water. Further, while epithermal radiography significantly improves the radiographs, defects may go undetected. As an alternative to neutron radiography, high energy gamma radiographs of TRIGA fuel elements have been taken using the same facility. The gamma spectrum emitted by the reactor core is sufficiently high in energy that very good radiographs may be obtained with this technique. These radiographs show excellent detail for the internal structure of the TRIGA fuel. (author)

  18. Method for disposing of radioactive graphite and silicon carbide in graphite fuel elements

    SciTech Connect

    Gay, R.L.

    1995-09-12

    Method is described for destroying radioactive graphite and silicon carbide in fuel elements containing small spheres of uranium oxide coated with silicon carbide in a graphite matrix, by treating the graphite fuel elements in a molten salt bath in the presence of air, the salt bath comprising molten sodium-based salts such as sodium carbonate and a small amount of sodium sulfate as catalyst, or calcium-based salts such as calcium chloride and a small amount of calcium sulfate as catalyst, while maintaining the salt bath in a temperature range of about 950 to about 1,100 C. As a further feature of the invention, large radioactive graphite fuel elements, e.g. of the above composition, can be processed to oxidize the graphite and silicon carbide, by introducing the fuel element into a reaction vessel having downwardly and inwardly sloping sides, the fuel element being of a size such that it is supported in the vessel at a point above the molten salt bath therein. Air is bubbled through the bath, causing it to expand and wash the bottom of the fuel element to cause reaction and destruction of the fuel element as it gradually disintegrates and falls into the molten bath. 4 figs.

  19. Radioactive Fission Product Release from Defective Light Water Reactor Fuel Elements

    SciTech Connect

    Konyashov, Vadim V.; Krasnov, Alexander M.

    2002-04-15

    Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. An approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)

  20. Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

    NASA Technical Reports Server (NTRS)

    Klann, P. G.; Lantz, E.

    1973-01-01

    A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

  1. Analysis of cocked fuel elements in the AFRRI TRIGA Mark-F reactor

    SciTech Connect

    Sholtis, Joseph A. Jr.

    1982-07-01

    The Armed Forces Radiobiology Research Institute (AFRRI) TRIGA Mark-F pulsing reactor has experienced eight cocked fuel elements during the period 5 November 1974 through 17 February 1982. Although there are no adverse health and safety consequences associated with their occurrence and there is no credible potential for system damage, cocked TRIGA fuel elements do cause inconvenience to the reactor staff and a temporary delay in operations. This paper presents the history of cocked TRIGA fuel elements at AFRRI, discusses possible mechanisms for their occurrence, and outlines a plan to isolate and ultimately determine their actual cause.

  2. Inhalation of U aerosols from UO2 fuel element fabrication.

    PubMed

    Schieferdecker, H; Dilger, H; Doerfel, H; Rudolph, W; Anton, R

    1985-01-01

    Publication No. 30 of the International Commission on Radiological Protection (ICRP) assigns the uranium oxides UO2 and U3O8 to transportability class Y, i.e. the half-life of these compounds in the lungs is about 500 days. This assignment seemed not to be in accordance with our experience resulting from incorporation surveillance during UO2 fuel element fabrication. Persons who worked in atmospheres containing UO2 aerosols with activity concentrations significantly above the derived air concentrations (DAC) for class Y U showed much lower activity in the lungs than would be expected according to the ICRP. To understand this discrepancy, aerosol concentrations and aerosol particle-size distributions at work places with the possibility of UO2 incorporation, the activity of urine and feces and the lung activity of persons working at these places were measured in an investigation program. The results are only consistent with the ICRP lung model if one uses a measured biological half-life in the lungs of 109 days and a measured AMAD of 8.2 micron instead of the ICRP standard assumptions of 500 days and 1.0 micron, respectively. ICRP Publication No. 30 recommends application of specific parameters for health physics instead of standard model values. For the special conditions in our UO2 fuel fabrication plant we therefore derive limits of air concentrations, lung activities and fecal and urinary activity concentrations by applying our measured particle-size and lung-retention parameters to the ICRP model. Our special derived limits in comparison to class Y limits for U after ICRP Publication No. 30 for a 1-micron AMAD and 500-day half-life (in brackets) are: (a) annual limit of intake: 6 X 10(4) Bq/y (1 X 10(3) Bq/y); (b) derived air concentration: 20 Bq/m3 (0.6 Bq/m3); (c) derived lung activity: 1.6 X 10(3) Bq; (d) derived fecal activity: 14 Bq/day; and (e) derived urine activity: 8.9 Bq/day. The committed dose equivalents calculated from our measured data and from our

  3. Accelerator-driven transmutation of spent fuel elements

    DOEpatents

    Venneri, Francesco; Williamson, Mark A.; Li, Ning

    2002-01-01

    An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

  4. Performance and fuel-cycle cost analysis of one JANUS 30 conceptual design for several fuel-element-design options

    SciTech Connect

    Nurdin, M.; Matos, J.E.; Freese, K.E.

    1982-01-01

    The performance and fuel cycle costs for a 25 MW, JANUS 30 reactor conceptual design by INTERATOM, Federal Republic of Germany, for BATAN, Republic of Indonesia have been studied using 19.75% enriched uranium in four fuel element design options. All of these fuel element designs have either been proposed by INTERATOM for various reactors or are currently in use with 93% enriched uranium in reactors in the Federal Republic of Germany. Aluminide, oxide, and silicide fuels were studied for selected designs using the range of uranium densities that are either currently qualified or are being developed and demonstrated internationally. To assess the long-term fuel adaptation strategy as well as the present fuel acceptance, reactor performance and annual fuel cycle costs were computed for seventeen cases based on a representative end-of-cycle excess reactivity and duty factor. In addition, a study was made to provide data for evaluating the trade-off between the increased safety associated with thicker cladding and the economic penalty due to increased fuel consumption.

  5. Experimental investigation of fuel evaporation in the vaporizing elements of combustion chambers

    NASA Technical Reports Server (NTRS)

    Vezhba, I.

    1979-01-01

    A description is given of the experimental apparatus and the methods used in the investigation of the degree of fuel (kerosene) evaporation in two types of vaporizing elements in combustion chambers. The results are presented as dependences of the degree of fuel evaporation on the factors which characterize the functioning of the vaporizing elements: the air surplus coefficient, the velocity of flow and temperature of the air at the entrance to the vaporizing element and the temperature of the wall of the vaporizing element.

  6. Pumped lithium loop test to evaluate advanced refractory metal alloys and simulated nuclear fuel elements

    NASA Technical Reports Server (NTRS)

    Brandenburf, G. P.; Hoffman, E. E.; Smith, J. P.

    1974-01-01

    The performance was determined of refractory metal alloys and uranium nitride fuel element specimens in flowing 1900F (1083C) lithium. The results demonstrate the suitability of the selected materials to perform satisfactorily from a chemical compatibility standpoint.

  7. Current status of U{sub 3}Si{sub 2} fuel element fabrication in Brazil

    SciTech Connect

    Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.; Souza, J.A.B

    2008-07-15

    IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U{sub 3}O{sub 8}-Al dispersion fuel plates with 2.3 gU/cm{sup 3}. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicide was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)

  8. A combined wet/dry sipping cell for investigating failed TRIGA fuel elements

    SciTech Connect

    Hammer, J.; Gallhammer, H.; Bock, H.

    1988-07-01

    Investigation for a failed TRIGA fuel element is performed with the help of a combined wet/dry sipping cell, which has been designed and fabricated at the Atominstitut Vienna. In this sipping cell a TRIGA fuel element can be studied for fission product release, both at normal and at elevated temperatures. This report describes the design features of the sipping cell and the fission product identification procedure with the help of a high purity Germanium detector and a multichannel analyzer.

  9. MECHANICALLY-JOINED PLATE-TYPE ALUMINUM-CLAD FUEL ELEMENT

    DOEpatents

    Erwin, J.H.

    1962-12-11

    A method of fabricating MTR-type fuel elements is described wherein dove- tailed joints are used to fasten fuel plates to supporting side members. The method comprises the steps of dove-tailing the lateral edges of the fuel plates, inserting the dove-tailed edges into corresponding recesses which are provided in a pair of supporting side members, and compressing the supporting side members in a direction so as to close the recesses onto the dove-tailed edges. (AEC)

  10. Aluminum hydroxide coating thickness measurements and brushing tests on K West Basin fuel elements

    SciTech Connect

    Pitner, A.L.

    1998-09-11

    Aluminum hydroxide coating thicknesses were measured on fuel elements stored in aluminum canisters in K West Basin using specially developed eddy current probes . The results were used to estimate coating inventories for MCO fuel,loading. Brushing tests successfully demonstrated the ability to remove the coating if deemed necessary prior to MCO loading.

  11. Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

    NASA Technical Reports Server (NTRS)

    Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

    2013-01-01

    In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

  12. Problems in developing bimodal space power and propulsion system fuel element

    SciTech Connect

    Nikolaev, Yu. V.; Gontar, A. S.; Zaznoba, V. A.; Parshin, N. Ya.; Ponomarev-Stepnoi, N. N.; Usov, V. A.

    1997-01-10

    The paper discusses design of a space nuclear power and propulsion system fuel element (PPFE) developed on the basis of an enhanced single-cell thermionic fuel element (TFE) of the 'TOPAZ-2' thermionic converter-reactor (TCR), and presents the PPFE performance for propulsion and power modes of operation. The choice of UC-TaC fuel composition is substantiated. Data on hydrogen effect on the PPFE output voltage are presented, design solutions are considered that allow to restrict hydrogen supply to an interelectrode gap (IEG). Long-term geometric stability of an emitter assembly is supported by calculated data.

  13. Apparatus for and method of monitoring for breached fuel elements

    DOEpatents

    Gross, Kenny C.; Strain, Robert V.

    1983-01-01

    This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus uses a separate bypass loop for conveying part of the reactor coolant away from the core, and at least three separate delayed-neutron detectors mounted proximate this detector loop. The detectors are spaced apart so that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the dealy time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector. At least two of these time components are determined during calibrated operation of the reactor. Thereafter during normal reactor operation, repeated comparisons are made by the method of regression approximation of the third time component for the best-fit line correlating measured delayed-neutron activity against activity that is approximated according to specific equations. The equations use these time-delay components and known parameter values of the fuel and of the part and emitting daughter isotopes.

  14. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    SciTech Connect

    Pope, M. A.; DeHart, M. D.; Morrell, S. R.; Jamison, R. K.; Nef, E. C.; Nigg, D. W.

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  15. Testing of sludge coating adhesiveness on fuel elements in 105-K west basin

    SciTech Connect

    Maassen, D.P., Fluor Daniel Hanford

    1997-03-11

    This report summarizes the results from the first sludge adherence tests performed in the 105-K West Basin on N Reactor fuel. The outside surface of the outer fuel elements were brushed, using stainless steel wire brushes, to test the adhesiveness of various types of sludge coatings to the cladding`s surface. The majority of the sludge was removed by the wire brushes in this test but different types of sludge were more adhesive than others. Particularly, an orange rust-like sludge coating that was just slightly more adherent to the fuel`s cladding than the majority of the sludge coatings and a thick white vertical strip sludge coating that was much more difficult to remove. The test demonstrated that all of the sludge could be removed from the outer fuel elements` surfaces if the need arises.

  16. Drying Results of K-Basin Fuel Element 6603M (Rune 5)

    SciTech Connect

    B.M. Oliver; G.A. Ritter; G.S. Klinger; J. Abrefah; L.R. Greenwood; P.J. MacFarlan; S.C. Marschman

    1999-09-24

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium spent nuclear fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fifth of those tests conducted on an N-Reactor outer fuel element (6603M) which had been stored underwater in the Hanford 100 Area K-West basin from 1983 until 1996. This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments which were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0. The test conditions and methodologies are given in Section 3.0. Inspections on the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0. Discussion of the results is given in Section 6.0.

  17. Method for recovering catalytic elements from fuel cell membrane electrode assemblies

    DOEpatents

    Shore, Lawrence; Matlin, Ramail; Heinz, Robert

    2012-06-26

    A method for recovering catalytic elements from a fuel cell membrane electrode assembly is provided. The method includes converting the membrane electrode assembly into a particulate material, wetting the particulate material, forming a slurry comprising the wetted particulate material and an acid leachate adapted to dissolve at least one of the catalytic elements into a soluble catalytic element salt, separating the slurry into a depleted particulate material and a supernatant containing the catalytic element salt, and washing the depleted particulate material to remove any catalytic element salt retained within pores in the depleted particulate material.

  18. Management of the spent fuel elements of the thorium high temperature reactor THTR-300

    SciTech Connect

    Quaassdorff, P.; Mielisch, M.; Dietrich, G.; Heske, M.; Jacobsen, W.

    1995-12-31

    In a world-wide unique campaign ca. 620,000 spent fuel elements of the thorium high temperature reactor THTR 300 which is being decommissioned, were being transferred within a short period of time to the Ahaus fuel element interim store (BZA) for interim storage. In order to optimize the technical and logistic procedures as part of the pre-decommissioning operation in 1992 and 1993, 42,000 fuel elements which had already been removed from the reactor core were transferred to Ahaus in transport and storage casks of the CASTOR THTR/AVR type that have been specially designed for this purpose. The experiences gained with loading, processing and transport of 20 transport and storage casks during this optimization and testing period led the team to expect a smooth management of the remaining fuel elements. In January 1994, the routine operation of the outward transfer commenced. Until mid-November 1994, 554,400 spent fuel elements were transferred outward into altogether 264 transport and storage casks of the CASTOR THTR/AVR type and transported to Ahaus for interim storage. This was followed by processing of another 21 transport and storage casks until April 1995, accommodating damaged fuel elements and special elements. The work mentioned above was performed by SFEAG Kernenergie GmbH, Essen, on behalf of the reactor operator Hochtemperatur-Kernkraftwerk GmbH, Hamm. The removal of the nuclear fuel from the thorium high temperature reactor THTR-300 marks the completion of the first part of the necessary actions for the decommissioning of the reactor (safe enclosure).

  19. Nondestructive examination of 51 fuel and reflector elements from Fort St. Vrain Core Segment 1

    SciTech Connect

    Miller, C.M.; Saurwein, J.J.

    1980-12-01

    Fifty-one fuel and reflector elements irradiated in core segment 1 of the Fort St. Vrain High-Temperature Gas-Cooled Reactor (HTGR) were inspected dimensionally and visually in the Hot Service Facility at Fort St. Vrain in July 1979. Time- and volume-averaged graphite temperatures for the examined fuel elements ranged from approx. 400/sup 0/ to 750/sup 0/C. Fast neutron fluences varied from approx. 0.3 x 10/sup 25/ n/m/sup 2/ to 1.0 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/. Nearly all of the examined elements shrank in both axial and radial dimensions. The measured data were compared with strain and bow predictions obtained from SURVEY/STRESS, a computer code that employs viscoelastic beam theory to calculate stresses and deformations in HTGR fuel elements.

  20. Molten tin reprocessing of spent nuclear fuel elements

    DOEpatents

    Heckman, Richard A.

    1983-01-01

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

  1. Uranium chloride extraction of transuranium elements from LWR fuel

    DOEpatents

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

    1992-08-25

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

  2. Uranium chloride extraction of transuranium elements from LWR fuel

    DOEpatents

    Miller, William E.; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Pierce, R. Dean

    1992-01-01

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

  3. Magnesium transport extraction of transuranium elements from LWR fuel

    DOEpatents

    Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.; Pierce, R. Dean

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.

  4. Apparatus for and method of monitoring for breached fuel elements

    DOEpatents

    Gross, K.C.; Strain, R.V.

    1981-04-28

    This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus uses a separate bypass loop for conveying part of the reactor coolant away from the core, and at least three separate delayed-neutron detectors mounted proximate this detector loop. The detectors are spaced apart so that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the delay time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector.

  5. AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

    SciTech Connect

    D. D. Keiser; J. I. Cole

    2007-09-01

    Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. This temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.

  6. Salt transport extraction of transuranium elements from lwr fuel

    DOEpatents

    Pierce, R. Dean; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750.degree. C. to about 850.degree. C. to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including Mg Cl.sub.2 to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy.

  7. Salt transport extraction of transuranium elements from LWR fuel

    DOEpatents

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

    1992-11-03

    A process is described for separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl[sub 2] and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750 C to about 850 C to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl[sub 2] having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO[sub 2]. The Ca metal and CaCl[sub 2] is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including MgCl[sub 2] to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy. 2 figs.

  8. Accuracy of trace element determinations in alternate fuels

    NASA Technical Reports Server (NTRS)

    Greenbauer-Seng, L. A.

    1980-01-01

    A review of the techniques used at Lewis Research Center (LeRC) in trace metals analysis is presented, including the results of Atomic Absorption Spectrometry and DC Arc Emission Spectrometry of blank levels and recovery experiments for several metals. The design of an Interlaboratory Study conducted by LeRC is presented. Several factors were investigated, including: laboratory, analytical technique, fuel type, concentration, and ashing additive. Conclusions drawn from the statistical analysis will help direct research efforts toward those areas most responsible for the poor interlaboratory analytical results.

  9. PROCESS OF MAKING A NEUTRONIC REACTOR FUEL ELEMENT COMPOSITION

    DOEpatents

    Alter, H.W.; Davidson, J.K.; Miller, R.S.; Mewherter, J.L.

    1959-01-13

    A process is presented for making a ceramic-like material suitable for use as a nuclear fuel. The material consists of a solid solution of plutonium dioxide in uranium dioxide and is produced from a uranyl nitrate -plutonium nitrate solution containing uraniunm and plutonium in the desired ratio. The uranium and plutonium are first precipitated from the solution by addition of NH/ sub 4/OH and the dried precipitate is then calcined at 600 C in a hydrogen atmosphere to yield the desired solid solution of PuO/sub 2/ in UO/sub 2/.

  10. Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

    NASA Technical Reports Server (NTRS)

    Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

    2012-01-01

    With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

  11. Analytical Solution of Fick's Law of the TRISO-Coated Fuel Particles and Fuel Elements in Pebble-Bed High Temperature Gas-Cooled Reactors

    NASA Astrophysics Data System (ADS)

    Cao, Jian-Zhu; Fang, Chao; Sun, Li-Feng

    2011-05-01

    Two kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation. In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.

  12. Operational experience of ultrasonic sealing bolts for safeguard containment of multi-element bottles in British Nuclear Fuel`s THORP spent fuel storage ponds

    SciTech Connect

    Hatt, C.D.; Reynolds, A.F.; Jeffrey, A.; DeTourbet, P.; D`Agraives, B.; Toornvliet, J.; Wilt, B.

    1995-12-31

    Following verification of the presence of Light Water Reactor fuel stored in multi-element bottles (MEBs), in British Nuclear Fuel`s (BNFL), Thermal Oxide Reprocessing Plant (THORP) fuel storage pond by Euratom and the IAEA, one lid bolt is replaced by an Ultrasonic Sealing Bolt. This safeguards seal, developed by Euratom`s Joint Research Centre at Ispra, Italy, has been field tested at Sellafield over several years and applied.in volume since 1994. The use of sealing bolts and video surveillance provides dual containment/surveillance on the THORP storage ponds, and brings significant savings in time and hence cost to the operator at the annual inventory verification. Time savings of up to 80% are achievable compared to fuel verification using a collimated gamma detector.

  13. URANIUM OXIDE-CONTAINING FUEL ELEMENT COMPOSITION AND METHOD OF MAKING SAME

    DOEpatents

    Handwerk, J.H.; Noland, R.A.; Walker, D.E.

    1957-09-10

    In the past, bodies formed of a mixture of uranium dioxide and aluminum powder have been used in fuel elements; however, these mixtures were found not to be suitable when exposed to temperatures of about 600 deg C, because at such high temperatures the fuel elements were distorted. If uranosic oxide, U/sub 3/O/sub 8/, is substituted for UO/sub 2/, the mechanical properties are not impaired when these materials are used at about 600 deg C and no distortion takes place. The uranosic oxide and aluminum, both in powder form, are first mixed, and after a homogeneous mixture has been obtained, are shaped into fuel elements by extrusion at elevated temperature. Magnesium powder may be used in place of the aluminum.

  14. Metallographic examination of damaged N reactor spent nuclear fuel element SFEC5,4378

    SciTech Connect

    Marschman, S.C.; Pyecha, T.D.; Abrefah, J.

    1997-08-01

    N-Reactor spent nuclear fuel (SNF) is currently residing underwater in the K Basins at the Hanford site, in Richland, Washington. This report presents results of the metallographic examination of specimens cut from an SNF element (Mark IV-E) with breached cladding. The element had resided in the K-West (KW) Storage Basin for at least 10 years after it was discharged from the N-Reactor. The storage containers in the KW Basin were nominally closed, isolating the SNF elements from the open pool environment. Seven specimens from this Mark IV-E outer fuel element were examined using an optical metallograph. Included were two specimens that had been subjected to a conditioning process recommended by the Independent Technical Assessment Team, two specimens that had been subjected to a conditioning process recommended in the Integrated Process Strategy Report, and three that were in the as-received, as-cut condition. One of the as-received specimens had been cut from the damaged (or breached) end of the element. All other specimens were cut from the undamaged mid-region of the fuel element. The specimens were visually examined to (1) identify uranium hydride inclusions present in the uranium metal fuel, (2) measure the thickness of the oxide layer formed on the uranium edges and assess the apparent integrity and adhesion of the oxide layer, and (3) look for features in the microstructure that might provide an insight into the various corrosion processes that occurred during underwater storage in the KW Basin. These features included, but were not limited to, the integrity of the cladding and the fuel-to-cladding bond, obvious anomalies in the microstructure, excessive pitting or friability of the fuel matrix, and obvious anomalies in the distribution of uranium hydride or uranium carbide inclusions. Also, the observed metallographic features of the conditioned specimens were compared with those of the as-received (unconditioned) specimens. 11 refs., 93 figs., 2 tabs.

  15. Irradiation testing of full-sized, reduced-enrichment fuel elements

    SciTech Connect

    Snelgrove, J.L.; Copeland, G.L.

    1983-01-01

    The current status of the irradiation testing of full-sized, reduced-enrichment fuel elements and fuel rods under the US Reduced Enrichment Research and Test Reactor Program is reported. Being tested are UAl/sub x/-Al, U/sub 3/O/sub 8/-Al, U/sub 3/Si/sub 2/-Al, and U/sub 3/Si-Al dispersion fuels and UZrH/sub x/ (TRIGA) fuel at uranium densities in the fuel meat ranging from 1.7 to 6.0 Mg/m/sup 3/. Generally good performance has been experienced to date. Some preliminary results of postirradiation examinations are also included. A whole-core demonstration in the Oak Ridge Research Reactor is planned. Some details of this demonstration are provided.

  16. 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.

  17. 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.

  18. 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.

  19. Determination of trace elements in automotive fuels by filter furnace atomic absorption spectrometry

    NASA Astrophysics Data System (ADS)

    Anselmi, Anna; Tittarelli, Paolo; Katskov, Dmitri A.

    2002-03-01

    The determination of Cd, Cr, Cu, Pb and Ni was performed in gasoline and diesel fuel samples by electrothermal atomic absorption spectrometry using the Transverse Heated Filter Atomizer (THFA). Thermal conditions were experimentally defined for the investigated elements. The elements were analyzed without addition of chemical modifiers, using organometallic standards for the calibration. Forty-microliter samples were injected into the THFA. Gasoline samples were analyzed directly, while diesel fuel samples were diluted 1:4 with n-heptane. The following characteristic masses were obtained: 0.8 pg Cd, 6.4 pg Cr, 12 pg Cu, 17 pg Pb and 27 pg Ni. The limits of determination for gasoline samples were 0.13 μg/kg Cd, 0.4 μg/kg Cr, 0.9 μg/kg Cu, 1.5 μg/kg Pb and 2.5 μg/kg Ni. The corresponding limit of determination for diesel fuel samples was approximately four times higher for all elements. The element recovery was performed using the addition of organometallic compounds to gasoline and diesel fuel samples and was between 85 and 105% for all elements investigated.

  20. Theoretical studies of transient criticality of irradiated fuel elements

    SciTech Connect

    Barbry, F.; Bonhomme, C.; Hague, P.; Mather, D.J.; Shaw, P.M.

    1987-01-01

    The use of transport flasks containing irradiated fuel is a common event, and their movements are strictly regulated by the national competent authority in order that an acceptable level of control of radiation hazards be maintained. Nonetheless it has been considered prudent to quantify the consequences of a particular hypothetical accident involving a transport package. The particular accident examined assumed that recriticality occurs during the refilling of a flask, and for the Commissariat a l'Energie Atomique (CEA) scenario, for which flasks are transported dry, the hypothetical accident occurs as the flask is slowly lowered into a storage pond. An alternative UK scenario assumes that the flask is being refilled, following breach, by a high-pressure hose. Thus, the consequences of such an accident were estimated by developing computer codes, Chateau by the CEA and Sartemp by the UK Atomic Energy Authority (UKAEA). This and other results show that the hypothetical accident in which a transport flask is brought to critical by the reentry of water gives at most a relatively mild event. In view of the considerably unlikely circumstances and conservative aspects introduced, this result shows that such an accident can be safely contained.

  1. Which Elements Should be Recycled for a Comprehensive Fuel Cycle?

    SciTech Connect

    Steven Piet; Trond Bjornard; Brent Dixon; Dirk Gombert; Robert Hill; Chris Laws; Gretchen Matthern; David Shropshire; Roald Wigeland

    2007-09-01

    Uranium recovery can reduce the mass of waste and possibly the number of waste packages that require geologic disposal. Separated uranium can be managed with the same method (near-surface burial) as used for the larger quantities of depleted uranium or recycled into new fuel. Recycle of all transuranics reduces long-term environmental burden, reduces heat load to repositories, extracts more energy from the original uranium ore, and may have significant proliferation resistance and physical security advantages. Recovery of short-lived fission products cesium and strontium can allow them to decay to low-level waste in facilities tailored to that need, rather than geologic disposal. This could also reduce the number and cost of waste packages requiring geologic disposal. These savings are offset by costs for separation, recycle, and storage systems. Recovery of technetium-99 and iodine-129 can allow them to be sent to geologic disposal in improved waste forms. Such separation avoids contamination of the other products (uranium) and waste (cesium-strontium) streams with long-lived radioisotopes so the material might be disposed as low-level waste. Transmutation of technetium and iodine is a possible future alternative.

  2. Device for the disengagement of a nuclear reactor fuel element from an articulated finger grapnel and method of using same

    SciTech Connect

    Chollet, F.

    1984-01-17

    Device for the underwater disengagement of a nuclear reactor fuel element from a grapnel with at least two articulated fingers. The device is designed to be placed on the end of a duct for positioning fuel elements and includes jacks for adjusting the relative positions of the device and the grapnel-fuel element unit and for maintaining these positions, further jacks for unfastening the fingers of the grapnel from the body of the grapnel and still further, jacks for tilting the fingers of the grapnel so as to enable the fingers to release their hold on the fuel element.

  3. CERAMIC FUEL ELEMENT MATERIAL FOR A NEUTRONIC REACTOR AND METHOD OF FABRICATING SAME

    DOEpatents

    Duckworth, W.H.

    1957-12-01

    This patent relates to ceramic composition, and to neutronic reactor fuel elements formed therefrom. These ceramic elements have high density and excellent strength characteristics and are formed by conventional ceramic casting and sintering at a temperature of about 2700 deg F in a nitrogen atmosphere. The composition consists of silicon carbide, silicon, uranium oxide and a very small percentage of molybdenum. Compositions containing molybdenum are markedly stronger than those lacking this ingredient.

  4. Proposed modification of an instrumented TRIGA fuel element so that it may be handled with a standard TRIGA fuel handling tool

    SciTech Connect

    Doane, Harry J.

    1992-07-01

    Instrumented fuel elements whose thermocouples are no longer functional are still a useful source of reactor fuel. Their usefulness is hampered somewhat by the extension tubing that must extend above water level to keep the thermocouple extension leads dry and to keep pool water from interacting with the gas tight lead seal which is made below the lower coupling in the extension tubing. This facility proposes to modify an instrumented TRIGA fuel element by removing the extension tubing at the lower coupling and attaching to it a top end fixture that is normally supplied with a standard TRIGA fuel element. This would then allow movement of the modified fuel element with a standard TRIGA fuel handling tool. This paper will present the considerations involved in performing this modification and the presenter will solicit any useful information that might be contributed by attendees of the TRIGA Owners' Conference. (author)

  5. Drying results of K-Basin fuel element 3128W (run 2)

    SciTech Connect

    Abrefah, J.; Klinger, G.S.; Oliver, B.M.; Marshman, S.C.; MacFarlan, P.J.; Ritter, G.A.; Flament, T.A.

    1998-07-01

    An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-East Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of N-Reactor spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from an open K-East canister (3128W) during the first fuel selection campaign conducted in 1995, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. Although it was judged to be breached during in-basin (i.e., K-Basin) examinations, visual inspection of this fuel element in the hot cell indicated that it was likely intact. Some scratches on the coating covering the cladding were identified before the furnace test. The drying test was conducted in the Whole Element Furnace Testing System located in G-Cell within the PTL. This test system is composed of three basic systems: the in-cell furnace equipment, the system gas loop, and the analytical instrument package. Element 3128W was subjected to the drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step. Results of the Pressure Rise and Gas Evolution Tests suggest that most of the free water in the system was released during the extended CVD cycle (68 hr versus 8 hr for the first run). An additional {approximately}0.34 g of water was released during the subsequent HVD phase, characterized by multiple water release peaks, with a principle peak at {approximately}180 C. This additional water is attributed to decomposition of a uranium hydrate (UO{sub 4}{center_dot}4H{sub 2}O/UO{sub 4}{center_dot}2H{sub 2}O) coating that was observed to be covering the surface

  6. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    SciTech Connect

    Not Available

    1990-03-01

    This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs.

  7. Conceptual design report for the mechanical disassembly of Fort St. Vrain fuel elements

    SciTech Connect

    Lord, D.L.; Wadsworth, D.C.; Sekot, J.P.; Skinner, K.L.

    1993-04-01

    A conceptual design study was prepared that: (1) reviewed the operations necessary to perform the mechanical disassembly of Fort St. Vrain fuel elements; (2) contained a description and survey of equipment capable of performing the necessary functions; and (3) performed a tradeoff study for determining the preferred concepts and equipment specifications. A preferred system was recommended and engineering specifications for this system were developed.

  8. Fuel-element failures in Hanford single-pass reactors 1944--1971

    SciTech Connect

    Gydesen, S.P.

    1993-07-01

    The primary objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emissions since 1944 from the US Department of Energy`s (DOE) Hanford Site near Richland, Washington. To estimate the doses, the staff of the Source Terms Task use operating information from historical documents to approximate the radioactive emissions. One source of radioactive emissions to the Columbia River came from leaks in the aluminum cladding of the uranium metal fuel elements in single-pass reactors. The purpose of this letter report is to provide photocopies of the documents that recorded these failures. The data from these documents will be used by the Source Terms Task to determine the contribution of single-pass reactor fuel-element failures to the radioactivity of the reactor effluent from 1944 through 1971. Each referenced fuel-element failure occurring in the Hanford single-pass reactors is addressed. The first recorded failure was in 1948, the last in 1970. No records of fuel-element failures were found in documents prior to 1948. Data on the approximately 2000 failures which occurred during the 28 years (1944--1971) of Hanford single-pass reactor operations are provided in this report.

  9. Review of Rover fuel element protective coating development at Los Alamos

    NASA Technical Reports Server (NTRS)

    Wallace, Terry C.

    1991-01-01

    The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

  10. An Expert System to Analyze Homogeneity in Fuel Element Plates for Research Reactors

    SciTech Connect

    Tolosa, S.C.; Marajofsky, A.

    2004-10-06

    In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up. This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to x-ray images. These images are generated when the x-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized x-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate.

  11. 2-D Time-Dependent Fuel Element, Thermal Analysis Code System.

    2001-09-24

    Version 00 WREM-TOODEE2 is a two dimensional, time-dependent, fuel-element thermal analysis program. Its primary purpose is to evaluate fuel-element thermal response during post-LOCA refill and reflood in a pressurized water reactor (PWR). TOODEE2 calculations are carried out in a two-dimensional mesh region defined in slab or cylindrical geometry by orthogonal grid lines. Coordinates which form order pairs are labeled x-y in slab geometry, and those in cylindrical geometry are labeled r-z for the axisymmetric casemore » and r-theta for the polar case. Conduction and radiation are the only heat transfer mechanisms assumed within the boundaries of the mesh region. Convective and boiling heat transfer mechanisms are assumed at the boundaries. The program numerically solves the two-dimensional, time-dependent, heat conduction equation within the mesh region. KEYWORDS: FUEL MANAGEMENT; HEAT TRANSFER; LOCA; PWR« less

  12. Chemical aspects of pellet-cladding interaction in light water reactor fuel elements

    SciTech Connect

    Olander, D.R.

    1982-01-01

    In contrast to the extensive literature on the mechanical aspects of pellet-cladding interaction (PCI) in light water reactor fuel elements, the chemical features of this phenomenon are so poorly understood that there is still disagreement concerning the chemical agent responsible. Since the earliest work by Rosenbaum, Davies and Pon, laboratory and in-reactor experiments designed to elucidate the mechanism of PCI fuel rod failures have concentrated almost exclusively on iodine. The assumption that this is the reponsible chemical agent is contained in models of PCI which have been constructed for incorporation into fuel performance codes. The evidence implicating iodine is circumstantial, being based primarily upon the volatility and significant fission yield of this element and on the microstructural similarity of the failed Zircaloy specimens exposed to iodine in laboratory stress corrosion cracking (SCC) tests to cladding failures by PCI.

  13. Conceptual design report for handling Fort St. Vrain fuel element components

    SciTech Connect

    Gavalya, R.A.

    1993-09-01

    This report presents conceptual designs for containment of high-level wastes (HLW) and low-level wastes (LLW) that will result from disassembly of fuel elements from the High Temperature Gas-Cooled Reactor at the Fort St. Vrain nuclear power plant in Platteville, Colorado. Hexagonal fuel elements will enter the disassembly area as a HLW and exit as either as HLW or LLW. The HLW will consist of spent fuel compacts that have been removed from the hexagonal graphite block. Graphite dust and graphite particles produced during the disassembly process will also be routed to the container that will hold the HLW spent fuel compacts. The LLW will consist of the emptied graphite block. Three alternatives have been introduced for interim storage of the HLW containers after the spent fuel has been loaded. The three alternatives are: (a) store containers where fuel elements are currently being stored, (b) construct a new dry storage facility, and (c) employ Multi-Purpose Canisters (currently in conceptual design stage). Containment of the LLW graphite block will depend on several factors: (a) LLW classification, (b) radiation levels, and (c) volume-reducing technique (if used). Packaging may range from cardboard boxes for incinerable wastes to 55-ton cask inserts for remote-handled wastes. Before final designs for the containment of the HLW and LLW can be developed, several issues need to be addressed: (a) packing factor for fuel compacts in HLW container, (b) storage/disposal of loaded HLW containers, (c) characterization of the emptied graphite blocks, and (d) which technique for volume-reduction purposes (if any) will be used.

  14. Examination of the surface coating removed from K-East Basin fuel elements

    SciTech Connect

    Abrefah, J.; Marschman, S.C.; Jenson, E.D.

    1998-05-01

    This report provides the results of studies conducted on coatings discovered on the surfaces of some N-Reactor spent nuclear fuel (SNF) elements stored at the Hanford K-East Basin. These elements had been removed from the canisters and visually examined in-basin during FY 1996 as part of a series of characterization tests. The characterization tests are being performed to support the Integrated Process Strategy developed to package, dry, transport, and store the SNF in an interim storage facility on the Hanford site. Samples of coating materials were removed from K-East canister elements 2350E and 2540E, which had been sent, along with nine other elements, to the Postirradiation Testing Laboratory (327 Building) for further characterization following the in-basin examinations. These coating samples were evaluated by Pacific Northwest National Laboratory using various analytical methods. This report is part of the overall studies to determine the drying behavior of corrosion products associated with the K-Basin fuel elements. Altogether, five samples of coating materials were analyzed. These analyses suggest that hydration of the coating materials could be an additional source of moisture in the Multi-Canister Overpacks being used to contain the fuel for storage.

  15. Time-resolved and time-integrated radiography of fast reactor fuel elements

    SciTech Connect

    De Volpi, A.

    1981-01-01

    The fast-reactor safety program has some unusual requirements in radiography. Applications may be divided into two areas: time-resolved or time-integrated radiography. The fast-neutron hodoscope has supplied all recent time-resolved cineradiographic in-pile fuel-motion data, and various x-ray and photographic techniques have been used for out-of-pile experiments. Thick containers and the large number of radioactive fuel pins involved in safety research have been responsible for some nonconventional applications of time-integrated radiography of stationary objects. Hodoscopes record fuel-motion during transient experiments at the TREAT reactor in the United States and CABRI in France. Other special techniques have been under development for out-of-pile nondestructive radiography of fuel element subassemblies, including fast-neutron and gamma-ray tomographic methods.

  16. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  17. TRISO-Fuel Element Performance Modeling for the Hybrid LIFE Engine with Pu Fuel Blanket

    SciTech Connect

    DeMange, P; Marian, J; Caro, M; Caro, A

    2010-02-18

    A TRISO-coated fuel thermo-mechanical performance study is performed for the hybrid LIFE engine to test the viability of TRISO particles to achieve ultra-high burnup of a weapons-grade Pu blanket. Our methodology includes full elastic anisotropy, time and temperature varying material properties for all TRISO layers, and a procedure to remap the elastic solutions in order to achieve fast fluences up to 30 x 10{sup 25} n {center_dot} m{sup -2} (E > 0.18 MeV). In order to model fast fluences in the range of {approx} 7 {approx} 30 x 10{sup 25} n {center_dot} m{sup -2}, for which no data exist, careful scalings and extrapolations of the known TRISO material properties are carried out under a number of potential scenarios. A number of findings can be extracted from our study. First, failure of the internal pyrolytic carbon (PyC) layer occurs within the first two months of operation. Then, the particles behave as BISO-coated particles, with the internal pressure being withstood directly by the SiC layer. Later, after 1.6 years, the remaining PyC crumbles due to void swelling and the fuel particle becomes a single-SiC-layer particle. Unrestrained by the PyC layers, and at the temperatures and fluences in the LIFE engine, the SiC layer maintains reasonably-low tensile stresses until the end-of-life. Second, the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Obtaining more reliable measurements, especially at higher fluences, is an imperative for the fidelity of our models. Finally, varying the geometry of the TRISO-coated fuel particles results in little differences in the scope of fuel performance. The mechanical integrity of 2-cm graphite pebbles that act as fuel matrix has also been studied and it is concluded that they can reliable serve the entire LIFE burnup cycle without failure.

  18. Choices of canisters and elements for the first fuel and canister sludge shipment from K East Basin

    SciTech Connect

    Makenas, B.J.

    1996-03-22

    The K East Basin contains open-top canisters with up to fourteen N Reactor fuel assemblies distributed between the two barrels of each canister. Each fuel assembly generally consists of inner and outer concentric elements fabricated from uranium metal with zirconium alloy cladding. The canisters also contain varying amounts of accumulated sludge. Retrieval of sample fuel elements and associated sludge for examination is scheduled to occur in the near future. The purpose of this document is to specify particular canisters and elements of interest as candidate sources of fuel and sludge to be shipped to laboratories.

  19. Comparison of HEU and LEU Fuel Neutron Spectrum for ATR Fuel Element and ATR Flux-Trap Positions

    SciTech Connect

    G. S. Chang

    2008-10-01

    The Advanced Test Reactor (ATR) is a high power and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the high total core power and high neutron flux, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. An optimized low-enriched uranium (LEU) (U-10Mo) core conversion case, which can meet the project requirements, has been selected. However, LEU contains a significant quantity of high density U-238 (80.3 wt.%), which will harden the neutron spectrum in the core region. Based on the reference ATR HEU and the optimized LEU full core plate-by-plate (PBP) models, the present work investigates and compares the neutron spectra differences in the fuel element (FE), Northeast flux trap (NEFT), Southeast flux trap (SEFT), and East flux trap (EFT) positions. A detailed PBP MCNP ATR core model was developed and validated for fuel cycle burnup comparison analysis. The current ATR core with HEU U 235 enrichment of 93.0wt.% was used as the reference model. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, an optimized LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.330 mm (13 mil) and the U-235 enrichment of 19.7 wt.% was used to calculate the impact of the neutron spectrum in FE and FT positions. MCNP-calculated results show that the neutron spectrum in the LEU FE is slightly harder than in the HEU FE, as expected. However, when neutrons transport through water coolant and beryllium (Be), the neutrons are thermalized to an equilibrium neutron spectrum as a function of water volume fraction in the investigated FT positions. As a result, the neutron spectrum differences of the HEU and LEU in the NEFT, SEFT, and EFT are negligible. To demonstrate that the LEU core fuel cycle performance can meet the

  20. Release Fractions from Multi-Element Spent Fuel Casks Resulting from HEDD Attack

    SciTech Connect

    Luna, R. E.

    2006-07-01

    This paper provides a simple model for estimating the release of respirable aerosols resulting from an attack on a spent fuel cask using a high energy density device (HEDD). Two primary experiments have provided data on potential releases from spent fuel casks under HEDD attack. Sandia National Laboratories (SNL) conducted the first in the early 1980's and the second was sponsored by Gessellshaft fur Anlagen- and Reaktorsicherheit (GRS) in Germany and conducted in France in 1994. Both used surrogate spent fuel assemblies in real casks. The SNL experiments used un-pressurized fuel pin assemblies in a single element cask while the GRS tests used pressurized fuel pin assemblies in a 9-element cask. Data from the two test programs is reasonably consistent, given the differences in the experiments, but the use of the test data for prediction of releases resulting from HEDD attack requires a method for accounting for the effects of pin pressurization release and the ratio of pin plenum gas release to cask free volume (VR). To account for the effects of VR and to link the two data sources, a simple model has been developed that uses both the SNL data and the GRS data as well as recent test data on aerosols produced in experiments with single pellets subjected to HEDD effects conducted under the aegis of the International Consortium's Working Group on Sabotage of Transport and Storage Casks (WGSTSC). (authors)

  1. The development of fuel performance models at the European institute for transuranium elements

    NASA Astrophysics Data System (ADS)

    Lassmann, K.; Ronchi, C.; Small, G. J.

    1989-07-01

    The design and operational performance of fuel rods for nuclear power stations has been the subject of detailed experimental research for over thirty years. In the last two decades the continuous demands for greater economy in conjunction with more stringent safety criteria have led to an increasing reliance on computer simulations. Conditions within a fuel rod must be calculated both for normal operation and for proposed reactor faults. It has thus been necessary to build up a reliable, theoretical understanding of the intricate physical, mechanical and chemical processes occurring under a wide range of conditions to obtain a quantitative insight into the behaviour of the fuel. A prime requirement, which has also proved to be the most taxing, is to predict the conditions under which failure of the cladding might occur, particularly in fuel nearing the end of its useful life. In this paper the general requirements of a fuel performance code are discussed briefly and an account is given of the basic concepts of code construction. An overview is then given of recent progress at the European Institute for Transuranium Elements in the development of a fuel rod performance code for general application and of more detailed mechanistic models for fission product behaviour.

  2. Reduced Toxicity Fuel Satellite Propulsion System Including Catalytic Decomposing Element with Hydrogen Peroxide

    NASA Technical Reports Server (NTRS)

    Schneider, Steven J. (Inventor)

    2002-01-01

    A reduced toxicity fuel satellite propulsion system including a reduced toxicity propellant supply for consumption in an axial class thruster and an ACS class thruster. The system includes suitable valves and conduits for supplying the reduced toxicity propellant to the ACS decomposing element of an ACS thruster. The ACS decomposing element is operative to decompose the reduced toxicity propellant into hot propulsive gases. In addition the system includes suitable valves and conduits for supplying the reduced toxicity propellant to an axial decomposing element of the axial thruster. The axial decomposing element is operative to decompose the reduced toxicity propellant into hot gases. The system further includes suitable valves and conduits for supplying a second propellant to a combustion chamber of the axial thruster, whereby the hot gases and the second propellant auto-ignite and begin the combustion process for producing thrust.

  3. Radiotoxicity and Risk Reduction of TRU Elements from Spent Fuel by Transmutation in the Light Water Reactor

    SciTech Connect

    Necas, Vladimir; Sebian, Vladimir; Kociskova, Karolina; Darilek, Petr

    2005-05-24

    A conventional PWR of type VVER-440 operating in a sustainable advanced fuel cycle mode with complete recycling of TRU elements in an Inert Matrix Combined Fuel Assembly (IMC-FA) in the same reactor was investigated. A preliminary assessment with the differences between various nuclear fuel cycles in terms of the risk analysis and its indicators has been conducted. The results indicate that the sustainable advanced fuel cycle option can, for the same amount of energy generation, significantly reduces both the amounts and radiotoxicity of the spent nuclear fuel in comparison with the conventional once-through UO2 or MOX fuel cycles.

  4. Fusion option to dispose of spent nuclear fuel and transuranic elements

    SciTech Connect

    Gohar, Y.

    2000-02-10

    The fusion option is examined to solve the disposition problems of the spent nuclear fuel and the transuranic elements. The analysis of this report shows that the top rated solution, the elimination of the transuranic elements and the long-lived fission products, can be achieved in a fusion reactor. A 167 MW of fusion power from a D-T plasma for sixty years with an availability factor of 0.75 can transmute all the transuranic elements and the long-lived fission products of the 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. The operating time can be reduced to thirty years with use of 334 MW of fusion power, a system study is needed to define the optimum time. In addition, the fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future. Fusion blankets with a liquid carrier for the transuranic elements can achieve a transmutation rate for the transuranic elements up to 80 kg/MW.y of fusion power with k{sub eff} of 0.98. In addition, the liquid blankets have several advantages relative to the other blanket options. The energy from this transmutation is utilized to produce revenue for the system. Molten salt (Flibe) and lithium-lead eutectic are identified as the most promising liquids for this application, both materials are under development for future fusion blanket concepts. The Flibe molten salt with transuranic elements was developed and used successfully as nuclear fuel for the molten salt breeder reactor in the 1960's.

  5. Computer modeling of single-cell and multicell thermionic fuel elements

    SciTech Connect

    Dickinson, J.W.; Klein, A.C.

    1996-05-01

    Modeling efforts are undertaken to perform coupled thermal-hydraulic and thermionic analysis for both single-cell and multicell thermionic fuel elements (TFE). The analysis--and the resulting MCTFE computer code (multicell thermionic fuel element)--is a steady-state finite volume model specifically designed to analyze cylindrical TFEs. It employs an interactive successive overrelaxation solution technique to solve for the temperatures throughout the TFE and a coupled thermionic routine to determine the total TFE performance. The calculated results include temperature distributions in all regions of the TFE, axial interelectrode voltages and current densities, and total TFE electrical output parameters including power, current, and voltage. MCTFE-generated results compare experimental data from the single-cell Topaz-II-type TFE and multicell data from the General Atomics 3H5 TFE to benchmark the accuracy of the code methods.

  6. Graphite corrosion and hydrogen release from HTR fuel elements in Q-brine

    SciTech Connect

    Fachinger, J.; Zhang, Z.X.; Brodda, B.G.

    1995-12-31

    Industrial reprocessing for High Temperature Reactors (HTR) fuel elements has never been installed in Germany. The spent fuel elements are being considered for final disposal in a rock salt repository in the deep geologic underground. Safety analysis requires the assumption of an accidental water ingress into the repository, resulting in the formation of a concentrated salt solution with the typical composition of a quinary brine. After corrosive penetration of the container walls, the brine may finally contact the fuel elements directly and mobilize radionuclides. Duve et al. investigated the leaching of the fission products and actinides from HTR fuel elements in Q-brine. The mobilization of {sup 14}C by graphite corrosion is one of the last data bases required as a source term for the release estimation of radionuclides in the final safety analysis. The evaluation of the hydrogen release was prescribed by the licensing board, because an excessive gas pressure may affect the overall integrity of the geological barrier. {sup 14}C occurs as dissolved organic and inorganic compounds in the brine. The leaching rate or organic {sup 14}C decreases from about 80 Bq to 1 Bq. The amount of organic {sup 14}C decreases from about 80 Bq to 1 Bq during leaching. The release of inorganic {sup 14}C ceases within 4 months. About 100 ppm of the total {sup 14}C inventory was released during leaching. Gaseous {sup 14}C has never been detected. The gas formation is based on the radiolytic degradation of water, with a formation rate of 0.04 to 0.11 ml/d. Gas chromatographic analysis of the gas proved that hydrogen is the main component of the released gas. Tritium and {sup 85}Kr were detected as traces with radio gas chromatography.

  7. Aerothermal modeling program, phase 2. Element C: Fuel injector-air swirl characterization

    NASA Technical Reports Server (NTRS)

    Mostafa, A. A.; Mongia, H. C.; Mcdonnell, V. G.; Samuelsen, G. S.

    1986-01-01

    The main objectives of the NASA-sponsored Aerothermal Modeling Program, Phase 2--Element C, are experimental evaluation of the air swirler interaction with a fuel injector in a simulated combustor chamber, assessment of the current two-phase models, and verification of the improved spray evaporation/dispersion models. This experimental and numerical program consists of five major tasks. Brief descriptions of the five tasks are given.

  8. Selection of representative volume elements for pore-scale analysis of transport in fuel cell materials

    NASA Astrophysics Data System (ADS)

    Wargo, E. A.; Hanna, A. C.; Çeçen, A.; Kalidindi, S. R.; Kumbur, E. C.

    2012-01-01

    Pore-scale modeling has become a quite popular tool for evaluating the impact of material structure on fuel cell performance. However, the computational complexity of these models often limits simulations to analyze only a small volume of material, which is typically selected randomly from a much larger microstructure dataset. When considering the heterogeneous internal structure of fuel cell materials, it is highly unlikely that such a randomly selected volume (i.e., model domain) would adequately reflect the salient features of the material structure. The objective of this work is to utilize the recent advances in microstructure quantification to select small representative volume elements (RVEs) that accurately reflect the overall microstructure and transport properties of fuel cell materials. The micro-porous layer (MPL) in polymer electrolyte fuel cells is chosen for initial demonstration of the approach. Dual-beam focused ion beam scanning electron microscopy is utilized to obtain a 3-D structural dataset of the selected MPL sample. The RVEs are selected using the new approach of weighted sets of optimally selected statistical volume elements, and the key structure and transport metrics are evaluated using advanced microstructure algorithms developed in-house. Metric comparisons between the RVEs and the full dataset indicate that the RVEs selected by this approach offer a very good representation of the full dataset, albeit in a volume that is significantly smaller in spatial extent, therefore providing a computationally efficient and reliable model domain for pore-scale analyses.

  9. Space shuttle orbit maneuvering engine, reusable thrust chamber program. Task 6: Data dump hot fuel element investigation

    NASA Technical Reports Server (NTRS)

    Nurick, W. H.

    1974-01-01

    An evaluation of reusable thrust chambers for the space shuttle orbit maneuvering engine was conducted. Tests were conducted using subscale injector hot-fire procedures for the injector configurations designed for a regenerative cooled engine. The effect of operating conditions and fuel temperature on combustion chamber performance was determined. Specific objectives of the evaluation were to examine the optimum like-doublet element geometry for operation at conditions consistent with a fuel regeneratively cooled engine (hot fuel, 200 to 250 F) and the sensitivity of the triplet injector element to hot fuels.

  10. Fabrication of zero power reactor fuel elements containing /sup 233/U/sub 3/O/sub 8/ powder

    SciTech Connect

    Nicol, R G; Parrott, J R; Krichinsky, A M; Box, W D; Martin, C W; Whitson, W R

    1982-05-01

    Oak Ridge National Laboratory, under contract with Argonne National Laboratory, completed the fabrication of 1743 fuel elements for use in their Zero Power Reactor. The contract also included recovery of 20 kg of /sup 233/U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U/sub 3/O/sub 8/ powder (suitable for fuel) and subsequent charging, sealing, decontamination, and testing of the fuel elements (packets) preparatory to shipment. The nuclear safety, radiation exposures, and quality assurance aspects of the program are discussed.

  11. FINITE ELEMENT ANALYSIS OF THE BONDED COMPLIANT SEAL DESIGN – A NEW SEALING CONCEPT FOR USE IN PLANAR SOLID OXIDE FUEL CELLS

    SciTech Connect

    Koeppel, Brian J.; Weil, K. Scott

    2005-08-01

    A key issue in developing commercially viable planar solid oxide fuel cell stacks is appropriate seal design. We are currently developing an alternative approach to rigid and compressive seal designs that conceptually combines advantages of both techiques, including hermeticity, mechanical integrity, and minimization of interfacial stresses in either of the joint substrate materials, particulary the ceramic. The new seal relies on a plastically deformable metal seal; one that offers a quasi-dynamic mechanical response in that it is adherent to both sealing surfaces, i.e. non-sliding, but readily yields or deforms under thermally generated stresses, thereby mitigating the development of stresses in the adjacent ceramic and metal components even when a significant difference in thermal expansion exists between the two materials. Here we employ finite element modeling to assess the potential thermal cycling performance of this design, specifically as it pertains to sealing components with vastly different thermal expansion properties.

  12. Information on the evolution of severe LWR fuel element damage obtained in the CORA program

    NASA Astrophysics Data System (ADS)

    Schanz, G.; Hagen, S.; Hofmann, P.; Schumacher, G.; Sepold, L.

    1992-06-01

    In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400°C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4C), which are leading to extensive low-temperature melt formation around 1200°C. Interrelations between those basic phenomena, resulting for example in cladding deformation ("flowering") and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ("quenching") are determining the evolution paths of fuel element destruction, which are to be identified. A further important task is the abstraction from mechanistic and microstructural details in order to get a rough classification of damage regimes (temperature and extent), a practicable analytical treatment of the materials behaviour, and a basis for decisions in accident mitigation and management procedures.

  13. Selenium and Other Elements in Water and Adjacent Rock and Sediment of Toll Gate Creek, Aurora, Arapahoe County, Colorado, December 2003 through March 2004

    USGS Publications Warehouse

    Herring, J.R.; Walton-Day, Katherine

    2007-01-01

    Streamwater and solid samples (rock, unconsolidated sediment, stream sediment, and efflorescent material) in the Toll Gate Creek watershed, Colorado, were collected and analyzed for major and trace elements to determine trace-element concentrations and stream loads from December 2003 through March 2004, a period of seasonally low flow. Special emphasis was given to selenium (Se) concentrations because historic Se concentrations exceeded current (2004) stream standards. The goal of the project was to assess the distribution of Se concentration and loads in Toll Gate Creek and to determine the potential for rock and unconsolidated sediment in the basin to be sources of Se to the streamwater. Streamwater samples and discharge measurements were collected during December 2003 and March 2004 along Toll Gate Creek and its two primary tributaries - West Toll Gate Creek and East Toll Gate Creek. During both sampling periods, discharge ranged from 2.5 liters per second to 138 liters per second in the watershed. Discharge was greater in March 2004 than December 2003, but both periods represent low flow in Toll Gate Creek, and results of this study should not be extended to periods of higher flow. Discharge decreased moving downstream in East Toll Gate Creek but increased moving downstream along West Toll Gate Creek and the main stem of Toll Gate Creek, indicating that these two streams gain flow from ground water. Se concentrations in streamwater samples ranged from 7 to 70 micrograms per liter, were elevated in the upstream-most samples, and were greater than the State stream standard of 4.6 micrograms per liter. Se loads ranged from 6 grams per day to 250 grams per day, decreased in a downstream direction along East Toll Gate Creek, and increased in a downstream direction along West Toll Gate Creek and Toll Gate Creek. The largest Se-load increases occurred between two sampling locations on West Toll Gate Creek during both sampling periods and between the two sampling

  14. Adjacent segment disease.

    PubMed

    Virk, Sohrab S; Niedermeier, Steven; Yu, Elizabeth; Khan, Safdar N

    2014-08-01

    EDUCATIONAL OBJECTIVES As a result of reading this article, physicians should be able to: 1. Understand the forces that predispose adjacent cervical segments to degeneration. 2. Understand the challenges of radiographic evaluation in the diagnosis of cervical and lumbar adjacent segment disease. 3. Describe the changes in biomechanical forces applied to adjacent segments of lumbar vertebrae with fusion. 4. Know the risk factors for adjacent segment disease in spinal fusion. Adjacent segment disease (ASD) is a broad term encompassing many complications of spinal fusion, including listhesis, instability, herniated nucleus pulposus, stenosis, hypertrophic facet arthritis, scoliosis, and vertebral compression fracture. The area of the cervical spine where most fusions occur (C3-C7) is adjacent to a highly mobile upper cervical region, and this contributes to the biomechanical stress put on the adjacent cervical segments postfusion. Studies have shown that after fusion surgery, there is increased load on adjacent segments. Definitive treatment of ASD is a topic of continuing research, but in general, treatment choices are dictated by patient age and degree of debilitation. Investigators have also studied the risk factors associated with spinal fusion that may predispose certain patients to ASD postfusion, and these data are invaluable for properly counseling patients considering spinal fusion surgery. Biomechanical studies have confirmed the added stress on adjacent segments in the cervical and lumbar spine. The diagnosis of cervical ASD is complicated given the imprecise correlation of radiographic and clinical findings. Although radiological and clinical diagnoses do not always correlate, radiographs and clinical examination dictate how a patient with prolonged pain is treated. Options for both cervical and lumbar spine ASD include fusion and/or decompression. Current studies are encouraging regarding the adoption of arthroplasty in spinal surgery, but more long

  15. Electrolyser and fuel cells, key elements for energy and life support

    NASA Astrophysics Data System (ADS)

    Bockstahler, Klaus; Funke, Helmut; Lucas, Joachim

    Both, Electrolyser and Fuel Cells are key elements for regenerative energy and life support systems. Electrolyser technology is originally intended for oxygen production in manned space habitats and in submarines, through splitting water into hydrogen and oxygen. Fuel cells serve for energy production through the reaction, triggered in the presence of an electrolyte, between a fuel and an oxidant. Now combining both technologies i.e. electrolyser and fuel cell makes it a Regenerative Fuel Cell System (RFCS). In charge mode, i.e. with energy supplied e.g. by solar cells, the electrolyser splits water into hydrogen and oxygen being stored in tanks. In discharge mode, when power is needed but no energy is available, the stored gases are converted in the fuel cell to generate electricity under the formation of water that is stored in tanks. Rerouting the water to the electrolyser makes it a closed-loop i.e. regenerative process. Different electrolyser and fuel cell technologies are being evolved. At Astrium emphasis is put on the development of an RFCS comprised of Fixed Alkaline Electrolyser (FAE) and Fuel Cell (AFC) as such technology offers a high electrical efficiency and thus reduced system weight, which is important in space applications. With increasing power demand and increasing discharge time an RFCS proves to be superior to batteries. Since the early technology development multiple design refinements were done at Astrium, funded by the European Space Agency ESA and the German National Agency DLR as well as based on company internal R and T funding. Today a complete RFCS energy system breadboard is established and the operational behavior of the system is being tested. In parallel the electrolyser itself is subject to design refinement and testing in terms of oxygen production in manned space habitats. In addition essential features and components for process monitoring and control are being developed. The present results and achievements and the dedicated

  16. Radionuclide Compositions and Total Activity of Spent MTR-HEU Fuel Elements of the IAN-R1 Research Reactor

    NASA Astrophysics Data System (ADS)

    Sarta, Josè A.; Castiblanco, Luis A.

    2005-05-01

    With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energìas Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer of the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.

  17. Radionuclide Compositions and Total Activity of Spent MTR-HEU Fuel Elements of the IAN-R1 Research Reactor

    SciTech Connect

    Sarta, Jose A.; Castiblanco, Luis A

    2005-05-24

    With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energias Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer of the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.

  18. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR... the integrity of completed fuel pins (or rods). This item typically includes equipment for: (i)...

  19. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR... the integrity of completed fuel pins (or rods). This item typically includes equipment for: (i)...

  20. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR... the integrity of completed fuel pins (or rods). This item typically includes equipment for: (i)...

  1. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR... the integrity of completed fuel pins (or rods). This item typically includes equipment for: (i)...

  2. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR... the integrity of completed fuel pins (or rods). This item typically includes equipment for: (i)...

  3. Film bonded fuel cell interface configuration

    DOEpatents

    Kaufman, Arthur; Terry, Peter L.

    1985-01-01

    An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. A multi-layer arrangement for the interface provides bridging electrical contact with a hot-pressed resin filling the void space.

  4. Atomistic modeling of the interaction of cladding elements (Fe, Ni, Cr) with U-Zr fuel

    NASA Astrophysics Data System (ADS)

    Bozzolo, G.; Mosca, H. O.; Yacout, A. M.; Hofman, G. L.

    2011-07-01

    Atomistic simulations of U-Zr fuel and its interaction with Fe, Ni, and Cr using the BFS method for alloys are presented. Results for the γU-βZr solid solution are discussed, including the behavior of the lattice parameter and coefficient of thermal expansion as a function of concentration and temperature. Output from these calculations is used to study the surface structure of γU-βZr for different crystallographic orientations, determining the concentration profiles, surface energy, and segregation behavior. The analysis is completed with simulations of the deposition of Fe, Ni and Cr on U-Zr substrates with varying Zr concentration. All results are discussed and interpreted by means of the concepts of strain and chemical energy underlying the BFS method, thus obtaining a simple explanation for the observed Zr segregation and its influence in allowing for cladding elements diffusion into the U-Zr fuel.

  5. Selective Catalytic Oxidation of Hydrogen Sulfide to Elemental Sulfur from Coal-Derived Fuel Gases

    SciTech Connect

    Gardner, Todd H.; Berry, David A.; Lyons, K. David; Beer, Stephen K.; Monahan, Michael J.

    2001-11-06

    The development of low cost, highly efficient, desulfurization technology with integrated sulfur recovery remains a principle barrier issue for Vision 21 integrated gasification combined cycle (IGCC) power generation plants. In this plan, the U. S. Department of Energy will construct ultra-clean, modular, co-production IGCC power plants each with chemical products tailored to meet the demands of specific regional markets. The catalysts employed in these co-production modules, for example water-gas-shift and Fischer-Tropsch catalysts, are readily poisoned by hydrogen sulfide (H{sub 2}S), a sulfur contaminant, present in the coal-derived fuel gases. To prevent poisoning of these catalysts, the removal of H{sub 2}S down to the parts-per-billion level is necessary. Historically, research into the purification of coal-derived fuel gases has focused on dry technologies that offer the prospect of higher combined cycle efficiencies as well as improved thermal integration with co-production modules. Primarily, these concepts rely on a highly selective process separation step to remove low concentrations of H{sub 2}S present in the fuel gases and produce a concentrated stream of sulfur bearing effluent. This effluent must then undergo further processing to be converted to its final form, usually elemental sulfur. Ultimately, desulfurization of coal-derived fuel gases may cost as much as 15% of the total fixed capital investment (Chen et al., 1992). It is, therefore, desirable to develop new technology that can accomplish H{sub 2}S separation and direct conversion to elemental sulfur more efficiently and with a lower initial fixed capital investment.

  6. Development of Nano-Sulfide Sorbent for Efficient Removal of Elemental Mercury from Coal Combustion Fuel Gas.

    PubMed

    Li, Hailong; Zhu, Lei; Wang, Jun; Li, Liqing; Shih, Kaimin

    2016-09-01

    The surface area of zinc sulfide (ZnS) was successfully enlarged using nanostructure particles synthesized by a liquid-phase precipitation method. The ZnS with the highest surface area (named Nano-ZnS) of 196.1 m(2)·g(-1) was then used to remove gas-phase elemental mercury (Hg(0)) from simulated coal combustion fuel gas at relatively high temperatures (140 to 260 °C). The Nano-ZnS exhibited far greater Hg(0) adsorption capacity than the conventional bulk ZnS sorbent due to the abundance of surface sulfur sites, which have a high binding affinity for Hg(0). Hg(0) was first physically adsorbed on the sorbent surface and then reacted with the adjacent surface sulfur to form the most stable mercury compound, HgS, which was confirmed by X-ray photoelectron spectroscopy analysis and a temperature-programmed desorption test. At the optimal temperature of 180 °C, the equilibrium Hg(0) adsorption capacity of the Nano-ZnS (inlet Hg(0) concentration of 65.0 μg·m(-3)) was greater than 497.84 μg·g(-1). Compared with several commercial activated carbons used exclusively for gas-phase mercury removal, the Nano-ZnS was superior in both Hg(0) adsorption capacity and adsorption rate. With this excellent Hg(0) removal performance, noncarbon Nano-ZnS may prove to be an advantageous alternative to activated carbon for Hg(0) removal in power plants equipped with particulate matter control devices, while also offering a means of reusing fly ash as a valuable resource, for example as a concrete additive. PMID:27508312

  7. Two-dimensional steady-state and transient analysis of single-cell thermionic fuel elements

    SciTech Connect

    El-Genk, M.S.; Xue, H. . Inst. for Space Nuclear Power Studies)

    1994-10-01

    A two-dimensional transient model is developed to simulate steady-state and transient operations of single-cell thermionic fuel elements (TFEs). Model predictions are in good agreement with published data to within 4.5 and 5.5% for fission and electrically heated TFEs of the TOPAZ-II type, respectively. In addition, the results of a transient analysis simulating the startup of an electrically heated TFE, following a step function increase in thermal power, are in presented and discussed.

  8. Multi-cell thermionic fuel element for nuclear electric power and propulsion system

    NASA Astrophysics Data System (ADS)

    Nikolaev, Yuri V.; Gontar, Alexander S.; Eremin, Stanislav A.; Lapochkin, Nikolai V.; Andreev, Pavel V.; Zhabotinsky, Evgeny E.

    1999-01-01

    Conceptual problems of development of two-mode multi-cell thermionic fuel element (TFE) for nuclear electric power and propulsion system are considered. The results of analysis of the design and TFE output parameters are presented. It is shown that application of advanced high effective materials and technologies provides operating of the TFE in two modes: a) in nominal mode of power generation for power supply of spacecraft payload at operational orbit and b) in forced mode of power generation for power supply of electric thrusters under spacecraft orbit transfer from intermediate to operational one.

  9. Irradiation and examinations of the second group of thermionic fuel element insulators (UCA-2)

    NASA Astrophysics Data System (ADS)

    Lawrence, Leo A.; Ard, Kevin E.; Veca, Anthony R.; Giraldez, Emilio M.

    1991-01-01

    Thermionic fuel element sheaths, seal and intercell insulators, and end restraints were irradiated in a fast neutron spectrum and examined. Samples were irradiated at temperatures ranging from 1110 K to 1200 K to fast fluences from 3.4×1022 n/cm2 to 6.0×1022 n/cm2. Sample examinations included visual, photographic, dimensional, electrical resistance to temperatures of 1175 K, helium leak rates, and metallography. Examinations of the end restraints and intercell insulators, which were limited to visual and photographic examination, showed no adverse effects from the irradiation. Alumina and yttria have been identified as insulator materials which meet design requirements.

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

    SciTech Connect

    SCHWINKENDORF, K.N.

    2006-05-12

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

  11. Dose Rate Calculations of Spent MTR-HEU Fuel Elements of the IAN-R1 Research Reactor

    NASA Astrophysics Data System (ADS)

    Sarta Fuentes, Jose Antonio

    2005-04-01

    With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safetly, a program was established at the Instituto de Ciencias Nucleares y Energìas Alternativas (INEA). This program included training, acquisition of hardware and sofware, design and construction of a decay pool, transfer of the spent HEU fuel elements into the decay pool and his final transport to Savanah River in United States. In this paper are presented external dose rates which were calculated for a standard spent MTR-HEU fuel element of the IAN-R1 Research Reactor. The calculations take in consideration the activity due to contributions of fission, activation and actinides products for each relevant radionuclide present in a standard spent MTR-HEU fuel. The datas obtained were the base for the respective dosimetric evaluations in the transfering operations of fuel elements into the decay pool and for shielding calculations in designing of the decay pool.

  12. Processing of FRG mixed oxide fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

    SciTech Connect

    Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Tischer, H.E.

    1980-11-01

    The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment section of the agreement, FRG fuel spheres were recently sent for processing in the Department of Energy sponsored cold pilot plant for High-Temperature Gas-Cooled Reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles. These particles were in turn crushed and burned to recover the fuel-bearing kernels for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated the applicability of the US HTGR fuel treatment flowsheet to FRG fuel processing. 10 figures.

  13. Processing of FRG high-temperature gas-cooled reactor fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

    SciTech Connect

    Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Drake, R.N.

    1981-11-01

    The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects of gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment development section of the agreement, both FRG mixed uranium/ thorium and low-enriched uranium fuel spheres have been processed in the Department of Energy-sponsored cold pilot plant for high-temperature gas-cooled reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles suitable for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated certain modifications to the US HTGR fuel burining process necessary for FRG fuel treatment. Results of the tests will be used in the design of a US/FRG joint prototype headend facility for HTGR fuel.

  14. Development of Low-Cost Manufacturing Processes for Planar, Multilayer Solid Oxide Fuel Cell Elements

    SciTech Connect

    Scott Swartz; Matthew Seabaugh; William Dawson; Tim Armstrong; Harlan Anderson; John Lannutti

    2001-09-30

    This report summarizes the results of Phase II of this program, 'Low-Cost Manufacturing Of Multilayer Ceramic Fuel Cells'. The objective of the program is to develop advanced ceramic manufacturing technologies for making planar solid oxide fuel cell (SOFC) components that are more economical and reliable for a variety of applications. Phase II development work focused on three distinct manufacturing approaches (or tracks) for planar solid oxide fuel cell elements. Two development tracks, led by NexTech Materials and Oak Ridge National Laboratory, involved co-sintering of planar SOFC elements of cathode-supported and anode-supported variations. A third development track, led by the University of Missouri-Rolla, focused on a revolutionary approach for reducing operating temperature of SOFCs by using spin-coating to deposit ultra-thin, nano-crystalline YSZ electrolyte films. The work in Phase II was supported by characterization work at Ohio State University. The primary technical accomplishments within each of the three development tracks are summarized. Track 1--NexTech's targeted manufacturing process for planar SOFC elements involves tape casting of porous electrode substrates, colloidal-spray deposition of YSZ electrolyte films, co-sintering of bi-layer elements, and screen printing of opposite electrode coatings. The bulk of NexTech's work focused on making cathode-supported elements, although the processes developed at NexTech also were applied to the fabrication of anode-supported cells. Primary accomplishments within this track are summarized below: (1) Scale up of lanthanum strontium manganite (LSM) cathode powder production process; (2) Development and scale-up of tape casting methods for cathode and anode substrates; (3) Development of automated ultrasonic-spray process for depositing YSZ films; (4) Successful co-sintering of flat bi-layer elements (both cathode and anode supported); (5) Development of anode and cathode screen-printing processes; and (6

  15. DEVELOPMENT OF LOW-COST MANUFACTURING PROCESSES FOR PLANAR, MULTILAYER SOLID OXIDE FUEL CELL ELEMENTS

    SciTech Connect

    Scott Swartz; Matthew Seabaugh; William Dawson; Harlan Anderson; Tim Armstrong; Michael Cobb; Kirby Meacham; James Stephan; Russell Bennett; Bob Remick; Chuck Sishtla; Scott Barnett; John Lannutti

    2004-06-12

    This report summarizes the results of a four-year project, entitled, ''Low-Cost Manufacturing Of Multilayer Ceramic Fuel Cells'', jointly funded by the U.S. Department of Energy, the State of Ohio, and by project participants. The project was led by NexTech Materials, Ltd., with subcontracting support provided by University of Missouri-Rolla, Michael A. Cobb & Co., Advanced Materials Technologies, Inc., Edison Materials Technology Center, Gas Technology Institute, Northwestern University, and The Ohio State University. Oak Ridge National Laboratory, though not formally a subcontractor on the program, supported the effort with separate DOE funding. The objective of the program was to develop advanced manufacturing technologies for making solid oxide fuel cell components that are more economical and reliable for a variety of applications. The program was carried out in three phases. In the Phase I effort, several manufacturing approaches were considered and subjected to detailed assessments of manufacturability and development risk. Estimated manufacturing costs for 5-kW stacks were in the range of $139/kW to $179/kW. The risk assessment identified a number of technical issues that would need to be considered during development. Phase II development work focused on development of planar solid oxide fuel cell elements, using a number of ceramic manufacturing methods, including tape casting, colloidal-spray deposition, screen printing, spin-coating, and sintering. Several processes were successfully established for fabrication of anode-supported, thin-film electrolyte cells, with performance levels at or near the state-of-the-art. The work in Phase III involved scale-up of cell manufacturing methods, development of non-destructive evaluation methods, and comprehensive electrical and electrochemical testing of solid oxide fuel cell materials and components.

  16. Elemental balance of SRF production process: solid recovered fuel produced from municipal solid waste.

    PubMed

    Nasrullah, Muhammad; Vainikka, Pasi; Hannula, Janne; Hurme, Markku; Oinas, Pekka

    2016-01-01

    In the production of solid recovered fuel (SRF), certain waste components have excessive influence on the quality of product. The proportion of rubber, plastic (hard) and certain textiles was found to be critical as to the elemental quality of SRF. The mass flow of rubber, plastic (hard) and textiles (to certain extent, especially synthetic textile) components from input waste stream into the output streams of SRF production was found to play the decisive role in defining the elemental quality of SRF. This paper presents the mass flow of polluting and potentially toxic elements (PTEs) in SRF production. The SRF was produced from municipal solid waste (MSW) through mechanical treatment (MT). The results showed that of the total input chlorine content to process, 55% was found in the SRF and 30% in reject material. Of the total input arsenic content, 30% was found in the SRF and 45% in fine fraction. In case of cadmium, lead and mercury, of their total input content to the process, 62%, 38% and 30%, respectively, was found in the SRF. Among the components of MSW, rubber material was identified as potential source of chlorine, containing 8.0 wt.% of chlorine. Plastic (hard) and textile components contained 1.6 and 1.1. wt.% of chlorine, respectively. Plastic (hard) contained higher lead and cadmium content compared with other waste components, i.e. 500 mg kg(-1) and 9.0 mg kg(-1), respectively. PMID:26608898

  17. Multiphysics Simulations of the Complex 3D Geometry of the High Flux Isotope Reactor Fuel Elements Using COMSOL

    SciTech Connect

    Freels, James D; Jain, Prashant K

    2011-01-01

    A research and development project is ongoing to convert the currently operating High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory (ORNL) from highly-enriched Uranium (HEU U3O8) fuel to low-enriched Uranium (LEU U-10Mo) fuel. Because LEU HFIR-specific testing and experiments will be limited, COMSOL is chosen to provide the needed multiphysics simulation capability to validate against the HEU design data and calculations, and predict the performance of the LEU fuel for design and safety analyses. The focus of this paper is on the unique issues associated with COMSOL modeling of the 3D geometry, meshing, and solution of the HFIR fuel plate and assembled fuel elements. Two parallel paths of 3D model development are underway. The first path follows the traditional route through examination of all flow and heat transfer details using the Low-Reynolds number k-e turbulence model provided by COMSOL v4.2. The second path simplifies the fluid channel modeling by taking advantage of the wealth of knowledge provided by decades of design and safety analyses, data from experiments and tests, and HFIR operation. By simplifying the fluid channel, a significant level of complexity and computer resource requirements are reduced, while also expanding the level and type of analysis that can be performed with COMSOL. Comparison and confirmation of validity of the first (detailed) and second (simplified) 3D modeling paths with each other, and with available data, will enable an expanded level of analysis. The detailed model will be used to analyze hot-spots and other micro fuel behavior events. The simplified model will be used to analyze events such as routine heat-up and expansion of the entire fuel element, and flow blockage. Preliminary, coarse-mesh model results of the detailed individual fuel plate are presented. Examples of the solution for an entire fuel element consisting of multiple individual fuel plates produced by the simplified model are also presented.

  18. The analysis of chlorine with other elements of interest in waste oil/fuels by ICP-AES

    SciTech Connect

    Tsourides, D.

    1998-12-31

    It has been said that there are more chemical analysis performed on oil/fuels than any other material. The sensitivity, linearity, multi-element capability, and relative freedom from matrix effects of ICP-AES makes it particularly suitable for elemental analysis of these samples. However, until recently the routine analysis of Chlorine had not been possible by ICP-AES. The addition of the Halogen elements, particularly Chlorine, to ICP-AES analysis is of importance to several industries that burn waste oil as fuel. The recycling and disposal of waste oil is closely regulated by metal and halogen content in all developed countries. In some countries, waste oil containing more than 1,000 ppm of Chlorine is considered hazardous waste. However, used oil may be burned as a fuel if it meets certain allowable limits. The paper describes the procedures for chlorine analysis by Inductively Coupled Plasma Atomic Emission Spectroscopy.

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

    SciTech Connect

    TOFFER, H.

    2006-07-18

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel

  20. Internal flow measurements of the SSME fuel preburner injector element using real time neutron radiography

    NASA Technical Reports Server (NTRS)

    Lindsay, John T.; Elam, Sandy; Koblish, Ted; Lee, Phil; Mcauliffe, Dave

    1990-01-01

    Due to observations of unsteady flow in the Space Shuttle Main Engine fuel preburner injector element, several flow studies have been performed. Real time neutron radiography tests were recently completed. This technique provided real time images of MiL-c-7024 and Freon-22 flow through an aluminum liquid oxygen post model at three back pressures (0, 150, and 545 psig) and pressure drops up to 1000 psid. Separated flow appeared only while operating at back pressures of 0 and 150 psig. The behavior of separated flow was similar to that observed for water in a 3x acrylic model of the LOX post. On the average, separated flow appeared to reattach near the exit of the post when the ratio of pressure drop to supply pressure was about 0.75.

  1. Output power characteristics and performance of TOPAZ II Thermionic Fuel Element No. 24

    SciTech Connect

    Luchau, D.W.; Bruns, D.R.; Izhvanov, O.; Androsov, V.

    1996-03-01

    A final report on the output power characteristics and capabilities of single cell TOPAZ II Thermionic Fuel Element (TFE) No. 24 is presented. Thermal power tests were conducted for over 3000 hours to investigate converter performance under normal and adverse operating conditions. Experiments conducted include low power testing, high power testing, air introduction to the interelectrode gap, collector temperature optimization, thermal modeling, and output power characteristic measurements. During testing, no unexpected degradation in converter performance was observed. The TFE has been removed from the test stand and returned to Scientific Industrial Association {open_quote}{open_quote}LUCH{close_quote}{close_quote} for materials analysis and report. This research was conducted at the Thermionic System Evaluation Test (TSET) Facility at the New Mexico Engineering Research Institute (NMERI) as a part of the Topaz International Program (TIP) by the Air Force Phillips Laboratory (PL). {copyright} {ital 1996 American Institute of Physics.}

  2. Mechanistic modeling of Zircaloy deformation and fracture in fuel element analysis

    SciTech Connect

    Nichols, F.A.

    1985-06-01

    A review is given of the comprehensive model developed in the 1960's at the Bettis Atomic Power Laboratory to explain the creep of Zircaloy during neutron irradiation and applied to fuel element analysis and design. The in-pile softening observed at low stresses was hypothesized to be due to a combination of the growth-directed Roberts-Cottrell yielding creep originally proposed for ..cap alpha..-uranium and the formation of point defect loops preferentially on certain planes in response to the applied stress, with the second process being of relatively greater importance. The in-pile hardening observed at high stresses (or strain-rates) was proposed to be due to the cutting by dislocations of radiation-produced obstacles. In this stress (strain-rate) region, in-pile behavior was proposed to be identical to post-irradiation behavior. At intermediate stresses (strain-rates) a mechanism of radiation-enhanced climb around obstacles was suggested as being rate controlling. As the stress is decreased, the climb process becomes easier and the rate was then predicted to be controlled by glide at a flow stress characteristic of unirradiated, annealed material, where radiation-enhanced diffusion enabled climbing around the normal strain-hardening obstacles. At still lower stresses, this glide process became negligibly slow compared with the growth-connected creep mechanism which was presumed to operate independently. The overall scheme was shown to be good agreement with all the in-pile data then available and implemented into the computer analysis of fuel element behavior. 48 refs., 1 fig.

  3. Utilization of a finite element model to verify spent nuclear fuel storage rack welds

    SciTech Connect

    Nitzel, M.E.

    1998-07-01

    Elastic and plastic finite element analyses were performed for the inner tie block assembly of a 25 port fuel rack designed for installation at the Idaho National Engineering and Environmental Laboratory (INEEL) Idaho Chemical Processing Plant (ICPP). The model was specifically developed to verify the adequacy of certain welds joining components of the fuel storage rack assembly. The work scope for this task was limited to an investigation of the stress levels in the inner tie welds when the rack was subjected to seismic loads. Structural acceptance criteria used for the elastic calculations performed were as defined by the rack`s designer. Structural acceptance criteria used for the plastic calculations performed as part of this effort were as defined in Subsection NF and Appendix F of Section III of the ASME Boiler and Pressure Vessel Code. The results confirm that the welds joining the inner tie block to the surrounding rack structure meet the acceptance criteria. The analysis results verified that the inner tie block welds should be capable of transferring the expected seismic load without structural failure.

  4. Fabrication of ORNL Fuel Irradiated in the Peach Bottom Reactor and Postirradiation Examination of Recycle Test Elements 7 and 4

    SciTech Connect

    Long, Jr. E.L.

    2001-10-25

    Seven full-sized Peach Bottom Reactor. fuel elements were fabricated in a cooperative effort by Oak Ridge National Laboratory (ORNL) and Gulf General Atomic (GGA) as part of the National HTGR Fuel Recycle Development Program. These elements contain bonded fuel rods and loose beds of particles made from several combinations of fertile and fissile particles of interest for present and future use in the High-Temperature Gas-Cooled Reactor (HTGR). The portion of the fuel prepared for these elements by ORNL is described in detail in this report, and it is in conjunction with the GGA report (GA-10109) a complete fabrication description of the test. In addition, this report describes the results obtained to date from postirradiation examination of the first two elements removed from the Peach Bottom Reactor, RTE-7 and -4. The fuel examined had relatively low exposure, up to about 1.5 x 10{sup 21} neutrons/cm* fast (>0.18 MeV) fluence, compared with the peak anticipated HTGR fluence of 8.0 x 10{sup 21}, but it has performed well at this exposure. Dimensional data indicate greater irradiation shrinkage than expected from accelerated test data to higher exposures. This suggests that either the method of extrapolation of the higher exposure data back to low exposure is faulty, or the behavior of the coated particles in the neutron spectrum characteristic of the accelerated tests does not adequately represent the behavior in an HTGR spectrum.

  5. Disposition of Unirradiated Sodium Bonded EBR-II Driver Fuel Elements and HEU Scrap: Work Performed for FY 2007

    SciTech Connect

    Karen A Moore

    2007-04-01

    Specific surplus high enriched uranium (HEU) materials at the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) will be transferred to a designated off-site receiving facility. The DOE High Enriched Uranium Disposition Program Office (HDPO) will determine which materials, if any, will be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for nuclear reactors. These surplus HEU materials include approximately 7200 kg unirradiated sodium-bonded EBR-II driver fuel elements, and nearly 800 kg of HEU casting scrap from the process which formed various sodium-bonded fuels (including the EBR-II driver elements). Before the driver fuel can be packaged for shipment, the fuel elements will require removal of the sodium bond. The HEU scrap will also require repackaging in preparation for off-site transport. Preliminary work on this task was authorized by BWXT Y-12 on Nov 6, 2006 and performed in three areas: • Facility Modifications • Safety Documentation • Project Management

  6. 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

  7. 3D laser inspection of fuel assembly grid spacers for nuclear reactors based on diffractive optical elements

    NASA Astrophysics Data System (ADS)

    Finogenov, L. V.; Lemeshko, Yu A.; Zav'yalov, P. S.; Chugui, Yu V.

    2007-06-01

    Ensuring the safety and high operation reliability of nuclear reactors takes 100% inspection of geometrical parameters of fuel assemblies, which include the grid spacers performed as a cellular structure with fuel elements. The required grid spacer geometry of assembly in the transverse and longitudinal cross sections is extremely important for maintaining the necessary heat regime. A universal method for 3D grid spacer inspection using a diffractive optical element (DOE), which generates as the structural illumination a multiple-ring pattern on the inner surface of a grid spacer cell, is investigated. Using some DOEs one can inspect the nomenclature of all produced grids. A special objective has been developed for forming the inner surface cell image. The problems of diffractive elements synthesis, projecting optics calculation, adjusting methods as well as calibration of the experimental measuring system are considered. The algorithms for image processing for different constructive elements of grids (cell, channel hole, outer grid spacer rim) and the experimental results are presented.

  8. Finite element thermal stress analysis of solid oxide fuel cell cathode microstructures

    NASA Astrophysics Data System (ADS)

    Vaidya, Sushrut; Kim, Jeong-Ho

    2013-03-01

    Two-dimensional images of solid oxide fuel cell (SOFC) cathode microstructures (50:50 wt.% LSM:YSZ) are used to generate three-dimensional finite element (FE) models. An approximate, heuristic scheme is developed to derive a microstructure of 30:70 wt.% LSM:YSZ composition from the original, real microstructures. The derived model is validated by calculating three-phase boundary (TPB) and phase surface area densities by comparing with data available in the literature. Construction of such derived microstructures will provide insights on the effects of phase compositions on the mechanics of electrode structures. The heuristic scheme is not proposed as a replacement for rigorous approaches such as the random packing model, but rather as a simplified approach for deriving reasonably realistic microstructures of different compositions within a limited range of validity. The models are analyzed using finite elements to estimate thermal stresses and probability of failure using Weibull analysis. The effects of temperature-dependent material properties and phase volume fractions on probability of failure of the cathode are discussed.

  9. Hydrogen loops in existing reactors for testing fuel elements for nuclear propulsion

    NASA Astrophysics Data System (ADS)

    Olsen, Charles S.; Welland, Henry; Abraschoff, James; Thoms, Kenneth

    1993-01-01

    The Space Exploration Initiative (SEI) has revitalized interest in adapting nuclear energy for power and propulsion. Prior to the selection of a nuclear thermal propulsion (NTP) system, extensive testing of the various proposed concepts will be required. In today's environmental, safety and health culture, full size rocket engine tests as were done under the Rover/NERVA program will be extremely difficult and expensive to perform and meet NASA's schedules. A different test strategy uses a hydrogen loop in an existing reactor to test a wide variety of single elements or clusters of elements for fuel qualification. This approach is expected to reduce operating and capital costs and expedite the testing schedule. This paper examines the potential of performing subscale tests in a hydrogen loop in an existing reactor such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. The HFIR is expected to achieve power densities comparable to those achieved in ATR because of the 85 MWt power level and the high thermal and fast flux levels. The available length and diameter of the test region of FHIR are 60 cm and 10 cm whereas the available length and diameter of the test region of ATR are 120 cm and 12 cm respectively.

  10. Hydrogen loops in existing reactors for testing fuel elements for nuclear propulsion

    SciTech Connect

    Olsen, C.S.; Welland, H.; Abraschoff, J. ); Thoms, K. )

    1993-01-15

    The Space Exploration Initiative (SEI) has revitalized interest in adapting nuclear energy for power and propulsion. Prior to the selection of a nuclear thermal propulsion (NTP) system, extensive testing of the various proposed concepts will be required. In today's environmental, safety and health culture, full size rocket engine tests as were done under the Rover/NERVA program will be extremely difficult and expensive to perform and meet NASA's schedules. A different test strategy uses a hydrogen loop in an existing reactor to test a wide variety of single elements or clusters of elements for fuel qualification. This approach is expected to reduce operating and capital costs and expedite the testing schedule. This paper examines the potential of performing subscale tests in a hydrogen loop in an existing reactor such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. The HFIR is expected to achieve power densities comparable to those achieved in ATR because of the 85 MWt power level and the high thermal and fast flux levels. The available length and diameter of the test region of FHIR are 60 cm and 10 cm whereas the available length and diameter of the test region of ATR are 120 cm and 12 cm respectively.

  11. Fuel injection and mixing systems having piezoelectric elements and methods of using the same

    DOEpatents

    Mao, Chien-Pei; Short, John; Klemm, Jim; Abbott, Royce; Overman, Nick; Pack, Spencer; Winebrenner, Audra

    2011-12-13

    A fuel injection and mixing system is provided that is suitable for use with various types of fuel reformers. Preferably, the system includes a piezoelectric injector for delivering atomized fuel, a gas swirler, such as a steam swirler and/or an air swirler, a mixing chamber and a flow mixing device. The system utilizes ultrasonic vibrations to achieve fuel atomization. The fuel injection and mixing system can be used with a variety of fuel reformers and fuel cells, such as SOFC fuel cells.

  12. Process for making film-bonded fuel cell interfaces

    DOEpatents

    Kaufman, Arthur; Terry, Peter L.

    1990-07-03

    An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. A multi-layer arrangement for the interface provides bridging electrical contact with a hot-pressed resin filling the void space.

  13. Base materials and technologies to maintain long service life and efficiency of thermionic converters and thermionic fuel elements

    NASA Astrophysics Data System (ADS)

    Nikolaev, Yury V.; Yastrebkov, Anotoly A.; Gontar, Alexander S.; Lapochkin, Nikolay V.; Belousenko, Alexander P.; Tsetskhladze, David L.

    2001-02-01

    It became possible to produce thermionic converters and thermionic fuel elements having a long-term service life and high efficiency only after developing new materials and processes of their production and treatment. This report present the characteristic of the level (achieved at present) of the base materials and technologies used in the State RI of SIA ``Lutch'' when producing TIC and TFE. .

  14. 78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-03

    ...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to Regulatory Guide (RG) 2.3, ``Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide describes a method that the staff of the NRC considers acceptable for complying with the Commission's regulations concerning establishing and executing a quality assurance......

  15. 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.

  16. Finite-element procedure for calculating the three-dimensional inelastic bowing of fuel rods (AWBA development program)

    SciTech Connect

    Martin, S E

    1982-05-01

    An incremental finite element procedure is developed for calculating the in-pile lateral bowing of nuclear fuel rods. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the results of an axisymmetric stress analysis of the fuel rod. The effects which are taken into account in calculating the rod's lateral bowing include: (a) lateral, axial, and rotational motions and forces at the rod supports, (b) transverse gradients of temperature, fast-neutron flux, and fissioning rate, and (c) cladding circumferential wall thickness variation. The procedure developed in this report could be used to form the basis for a computer program to calculate the time-dependent bowing as a function of the fuel rod's operational and environmental history.

  17. COMBINING NEUTRAL AND ACIDIC EXTRACTANTS FOR RECOVERING TRANSURANIC ELEMENTS FROM NUCLEAR FUEL

    SciTech Connect

    Lumetta, Gregg J.; Neiner, Doinita; Sinkov, Sergey I.; Carter, Jennifer C.; Braley, Jenifer C.; Latesky, Stanley; Gelis, Artem V.; Tkac, Peter; Vandegrift, George F.

    2011-10-03

    We have been investigating a solvent extraction system that combines a neutral extractant--octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide (CMPO)--with an acidic extractant--bis(2-ethylhexyl)phosphoric acid (HDEHP)--to form a single process solvent for separating Am and Cm from the other components of irradiated nuclear fuel. It was originally hypothesized that the extraction chemistry of CMPO would dominate under conditions of high acidity (> 1 M HNO3), resulting in co-extraction of the transuranic and lanthanide elements into the organic phase. Contacting the loaded solvent with a solution of diethylenetriaminepentaacetate (DTPA) buffered with lactic or citric acid at pH {approx}3 to 4 would result in a condition in which the HDEHP chemistry dominates. Although the data somewhat support this hypothesis, it is clear that there are interactions between the two extractants such that they do not act independently in the extraction and stripping regimes. We report here studies directed at determining the nature and extent of interaction between CMPO and HDEHP, the synergistic behavior of CMPO and HDEHP in the extraction of americium and neodymium, and progress towards determining the thermodynamics of this extraction system. Neodymium and americium behave similarly in the combined solvent system, with a significant synergy between CMPO and HDEHP in the extraction of both of these trivalent elements from lactate-buffered DTPA solutions. In contrast, a much weaker synergistic behaviour is observed for europium. Thus, investigations into the fundamental chemistry involved in this system have focused on the neodymium extraction. The extraction of neodymium has been systematically investigated, individually varying the HDEHP concentration, the CMPO concentration, or the aqueous phase composition. Thermodynamic modeling of the neodymium extraction system has been initiated. Interactions between CMPO and HDEHP in the organic phase must be taken into account in

  18. Two-dimensional steady-state analysis of an electrically heated thermionic fuel element

    SciTech Connect

    Huimin Xue; El-Genk, M.S.; Paramonov, D. )

    1993-01-20

    A two-dimensional transient model of a single cell, long Thermionic Fuel Element (TFE) is developed and its predictions are compared with published calculations and experimental data on steady-state operation of electrically heated, TOPAZ-II type TFEs. The operation parameters of the TFE, such as axial distributions of the emitter temperature, emission current density, and the electrode voltage are calculated and discussed. Results show that despite the excellent agreement between the model predictions of the axial distribution of the emitter temperature, its predictions of the maximum emission current density was lower by about 17%. This difference is attributed primarily to the J-V characteristics in the model, which could be different than those of the TOPAZ-II TFE, hence additional data on the latter is needed. When compared with experimental data, the model predictions of the electric power output are in excellent agreement with the data at thermal power input of 3.5 kW or higher, but within 10% of the data at lower thermal power.

  19. On-line elemental analysis of fossil fuel process streams by inductively coupled plasma spectrometry

    SciTech Connect

    Chisholm, W.P.

    1995-06-01

    METC is continuing development of a real-time, multi-element plasma based spectrometer system for application to high temperature and high pressure fossil fuel process streams. Two versions are under consideration for development. One is an Inductively Coupled Plasma system that has been described previously, and the other is a high power microwave system. The ICP torch operates on a mixture of argon and helium with a conventional annular swirl flow plasma gas, no auxiliary gas, and a conventional sample stream injection through the base of the plasma plume. A new, demountable torch design comprising three ceramic sections allows bolts passing the length of the torch to compress a double O-ring seal. This improves the reliability of the torch. The microwave system will use the same data acquisition and reduction components as the ICP system; only the plasma source itself is different. It will operate with a 750-Watt, 2.45 gigahertz microwave generator. The plasma discharge will be contained within a narrow quartz tube one quarter wavelength from a shorted waveguide termination. The plasma source will be observed via fiber optics and a battery of computer controlled monochromators. To extract more information from the raw spectral data, a neural net computer program is being developed. This program will calculate analyte concentrations from data that includes analyte and interferant spectral emission intensity. Matrix effects and spectral overlaps can be treated more effectively by this method than by conventional spectral analysis.

  20. Applying Thermodynamics to Fossil Fuels: Heats of Combustion from Elemental Compositions.

    ERIC Educational Resources Information Center

    Lloyd, William G.; Davenport, Derek A.

    1980-01-01

    Discussed are the calculations of heats of combustions of some selected fossil fuel compounds such as some foreign shale oils and United States coals. Heating values for coal- and petroleum-derived fuel oils are also presented. (HM)

  1. SAFE: A computer code for the steady-state and transient thermal analysis of LMR fuel elements

    SciTech Connect

    Hayes, S.L.

    1993-12-01

    SAFE is a computer code developed for both the steady-state and transient thermal analysis of single LMR fuel elements. The code employs a two-dimensional control-volume based finite difference methodology with fully implicit time marching to calculate the temperatures throughout a fuel element and its associated coolant channel for both the steady-state and transient events. The code makes no structural calculations or predictions whatsoever. It does, however, accept as input structural parameters within the fuel such as the distributions of porosity and fuel composition, as well as heat generation, to allow a thermal analysis to be performed on a user-specified fuel structure. The code was developed with ease of use in mind. An interactive input file generator and material property correlations internal to the code are available to expedite analyses using SAFE. This report serves as a complete design description of the code as well as a user`s manual. A sample calculation made with SAFE is included to highlight some of the code`s features. Complete input and output files for the sample problem are provided.

  2. TRISO-fuel element thermo-mechanical performance modeling for the hybrid LIFE engine with Pu fuel blanket

    NASA Astrophysics Data System (ADS)

    DeMange, P.; Marian, J.; Caro, M.; Caro, A.

    2010-10-01

    A TRISO-coated fuel thermo-mechanical performance study is performed for the fusion-fission hybrid Laser Inertial Fusion Engine (LIFE) to test the viability of TRISO particles to achieve ultra-high burn-up of Pu or transuranic spent nuclear fuel blankets. Our methodology includes full elastic anisotropy, time and temperature varying material properties, and multilayer capabilities. In order to achieve fast fluences up to 30 × 10 25 n m -2 ( E > 0.18 MeV), judicious extrapolations across several orders of magnitude of existing material databases have been carried out. The results of our study indicate that failure of the pyrolytic carbon (PyC) layers occurs within the first 2 years of operation. The particles then behave as a single-SiC-layer particle and the SiC layer maintains reasonably-low tensile stresses until the end-of-life. It is also found that the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Conversely, varying the geometry of the TRISO-coated fuel particles results in little differences in terms of fuel performance.

  3. Overview of past and current activities on fuels for fast reactors at the Institute for Transuranium Elements

    NASA Astrophysics Data System (ADS)

    Fernandez, A.; McGinley, J.; Somers, J.; Walter, M.

    2009-07-01

    Nuclear energy has the potential to provide a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gas emissions. The renewal of interest in fast neutron spectra reactors to meet more ambitious sustainable development criteria (i.e., resource maximisation and waste minimisation), opens a favourable framework for R&D activities in this area. The Institute for Transuranium Elements has extensive experience in the fabrication, characterization and irradiation testing (Phénix, Dounreay, Rapsodie) of fast reactor fuels, in oxide, nitride and carbide forms. An overview of these past and current activities on fast reactor fuels is presented.

  4. Distribution and leaching characteristics of trace elements in ashes as a function of different waste fuels and incineration technologies.

    PubMed

    Saqib, Naeem; Bäckström, Mattias

    2015-10-01

    Impact of waste fuels (virgin/waste wood, mixed biofuel (peat, bark, wood chips) industrial, household, mixed waste fuel) and incineration technologies on partitioning and leaching behavior of trace elements has been investigated. Study included 4 grate fired and 9 fluidized boilers. Results showed that mixed waste incineration mostly caused increased transfer of trace elements to fly ash; particularly Pb/Zn. Waste wood incineration showed higher transfer of Cr, As and Zn to fly ash as compared to virgin wood. The possible reasons could be high input of trace element in waste fuel/change in volatilization behavior due to addition of certain waste fractions. The concentration of Cd and Zn increased in fly ash with incineration temperature. Total concentration in ashes decreased in order of Zn>Cu>Pb>Cr>Sb>As>Mo. The concentration levels of trace elements were mostly higher in fluidized boilers fly ashes as compared to grate boilers (especially for biofuel incineration). It might be attributed to high combustion efficiency due to pre-treatment of waste in fluidized boilers. Leaching results indicated that water soluble forms of elements in ashes were low with few exceptions. Concentration levels in ash and ash matrix properties (association of elements on ash particles) are crucial parameters affecting leaching. Leached amounts of Pb, Zn and Cr in >50% of fly ashes exceeded regulatory limit for disposal. 87% of chlorine in fly ashes washed out with water at the liquid to solid ratio 10 indicating excessive presence of alkali metal chlorides/alkaline earths. PMID:26456601

  5. Best Practices for Finite Element Analysis of Spent Nuclear Fuel Transfer, Storage, and Transportation Systems

    SciTech Connect

    Bajwa, Christopher S.; Piotter, Jason; Cuta, Judith M.; Adkins, Harold E.; Klymyshyn, Nicholas A.; Fort, James A.; Suffield, Sarah R.

    2010-08-11

    Storage casks and transportation packages for spent nuclear fuel (SNF) are designed to confine SNF in sealed canisters or casks, provide structural integrity during accidents, and remove decay through a storage or transportation overpack. The transfer, storage, and transportation of SNF in dry storage casks and transport packages is regulated under 10 CFR Part 72 and 10 CFR Part 71, respectively. Finite Element Analysis (FEA) is used with increasing frequency in Safety Analysis Reports and other regulatory technical evaluations related to SNF casks and packages and their associated systems. Advances in computing power have made increasingly sophisticated FEA models more feasible, and as a result, the need for careful review of such models has also increased. This paper identifies best practice recommendations that stem from recent NRC review experience. The scope covers issues common to all commercially available FEA software, and the recommendations are applicable to any FEA software package. Three specific topics are addressed: general FEA practices, issues specific to thermal analyses, and issues specific to structural analyses. General FEA practices covers appropriate documentation of the model and results, which is important for an efficient review process. The thermal analysis best practices are related to cask analysis for steady state conditions and transient scenarios. The structural analysis best practices are related to the analysis of casks and associated payload during standard handling and drop scenarios. The best practices described in this paper are intended to identify FEA modeling issues and provide insights that can help minimize associated uncertainties and errors, in order to facilitate the NRC licensing review process.

  6. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    NASA Technical Reports Server (NTRS)

    Walton, James T.

    1992-01-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code.

  7. Novel, low-cost separator plates and flow-field elements for use in PEM fuel cells

    SciTech Connect

    Edlund, D.J.

    1996-12-31

    PEM fuel cells offer promise for a wide range of applications including vehicular (e.g., automotive) and stationary power generation. The performance and cost targets that must be met for PEM technology to be commercially successful varies to some degree with the application. However, in general the cost of PEM fuel cell stacks must be reduced substantially if they are to see widespread use for electrical power generation. A significant contribution to the manufactured cost of PEM fuel cells is the machined carbon plates that traditionally serve as bipolar separator plates and flow-field elements. In addition, carbon separator plates are inherently brittle and suffer from breakage due to shock, vibration, and improper handling. This report describes a bifurcated separator device with low resistivity, low manufacturing cost, compact size and durability.

  8. Distinct element method analyses of fuel spheres in the PBMR core using PFC{sup 3D}

    SciTech Connect

    Polson, Alexander G.

    2004-07-01

    The Pebble Bed Modular Reactor, or PBMR, is a High Temperature Gas Reactor that contains a large number of graphite fuel spheres that circulate in its core. The dynamics of these spheres, combined with thermal contraction and expansion, causes various loading cases on the reactor structures. A Distinct Element Method, or DEM, as implemented in the Particle Flow Code in 3D, or PFC{sup 3D}, is used at PBMR (Pty) Ltd to model the fuel sphere dynamics in the reactor core. This paper presents a few exploratory studies where PFC{sup 3D} was used to investigate the interaction between fuel spheres and structural components in the PBMR, as well as the packing efficiency of the spheres in the core. (author)

  9. Radiation dose rates from commercial PWR and BWR spent fuel elements

    SciTech Connect

    Willingham, C.E.

    1981-10-01

    Data on measurements of gamma dose rates from commercial reactor spent fuel were collected, and documented calculated gamma dose rates were reviewed. As part of this study, the gamma dose rate from spent fuel was estimated, using computational techniques similar to previous investigations into this problem. Comparison of the measured and calculated dose rates provided a recommended dose rate in air versus distance curve for PWR spent fuel.

  10. 3D modeling of heat transfer and gas flow in a grooved ring fuel element for nuclear thermal propulsion

    NASA Astrophysics Data System (ADS)

    Barkett, Laura Ashley

    In the past, fuel elements with multiple axial coolant channels have been used in nuclear propulsion applications. A novel fuel element concept that reduces weight and increases efficiency uses a stack of grooved rings. Each fuel ring consists of a hole on the interior and grooves across the top face. Many grooved ring configurations have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel ring with a higher surface-area-to-volume ratio is ideal. When grooves are shallower and they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of fluid flow with those of heat transfer, the effects on the cooler gas flowing through the grooves of the hot, fissioning ring can be predicted. Models also show differences in velocities and temperatures after dense boundary nodes are applied. Parametric studies were done to show how a pressure drop across the length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the temperature distributions and pressure drops that result from the manipulation of various parameters, and the effects of model scaling was also investigated. The inverse Graetz numbers are plotted against Nusselt numbers, and the results of these values suggest that the gas quickly becomes fully developed, laminar flow, rather than constant turbulent conditions.

  11. Corrosion protected, multi-layer fuel cell interface

    DOEpatents

    Feigenbaum, Haim; Pudick, Sheldon; Wang, Chiu L.

    1986-01-01

    An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. The multi-layer configuration for the interface comprises a non-cupreous metal-coated metallic element to which is film-bonded a conductive layer by hot pressing a resin therebetween. The multi-layer arrangement provides bridging electrical contact.

  12. A Multi-Dimensional Heat Transfer Model of a Tie-Tube and Hexagonal Fuel Element for Nuclear Thermal Propulsion

    NASA Technical Reports Server (NTRS)

    Gomez, C. F.; Mireles, O. R.; Stewart, E.

    2016-01-01

    The Space Capable Cryogenic Thermal Engine (SCCTE) effort considers a nuclear thermal rocket design based around a Low-Enriched Uranium (LEU) design fission reactor. The reactor core is comprised of bundled hexagonal fuel elements that directly heat hydrogen for expansion in a thrust chamber and hexagonal tie-tubes that house zirconium hydride moderator mass for the purpose of thermalizing fast neutrons resulting from fission events. Created 3D steady state Hex fuel rod model with 1D flow channels. Hand Calculation were used to set up initial conditions for fluid flow. The Hex Fuel rod uses 1D flow paths to model the channels using empirical correlations for heat transfer in a pipe. Created a 2-D axisymmetric transient to steady state model using the CFD turbulent flow and Heat Transfer module in COMSOL. This model was developed to find and understand the hydrogen flow that might effect the thermal gradients axially and at the end of the tie tube where the flow turns and enters an annulus. The Hex fuel rod and Tie tube models were made based on requirements given to us by CSNR and the SCCTE team. The models helped simplify and understand the physics and assumptions. Using pipe correlations reduced the complexity of the 3-D fuel rod model and is numerically more stable and computationally more time-efficient compared to the CFD approach. The 2-D axisymmetric tie tube model can be used as a reference "Virtual test model" for comparing and improving 3-D Models.

  13. Acceptance testing of the eddy current probes for measurement of aluminum hydroxide coating thickness on K West Basin fuel elements

    SciTech Connect

    Pitner, A.L.

    1998-08-21

    During a recent visual inspection campaign of fuel elements stored in the K West Basin, it was noted that fuel elements contained in sealed aluminum canisters had a heavy translucent type coating on their surfaces (Pitner 1997a). Subsequent sampling of this coating in a hot cell (Pitner 1997b) and analysis of the material identified it as aluminum hydroxide. Because of the relatively high water content of this material, safety related concerns are raised with respect to long term storage of this fuel in Multi-Canister Overpacks (MCOs). A campaign in the basin is planned to demonstrate whether this coating can be removed by mechanical brushing (Bridges 1998). Part of this campaign involves before-and-after measurements of the coating thickness to determine the effectiveness of coating removal by the brushing machine. Measurements of the as-deposited coating thickness on multiple fuel elements are also expected to provide total coating inventory information needed for MCO safety evaluations. The measurement technique must be capable of measuring coating thicknesses on the order of several mils, with a measurement accuracy of 0.5 mil. Several different methods for quantitatively measuring these thin coatings were considered in selecting the most promising approach. Ultrasonic measurement was investigated, but it was determined that due to the thin coating depth and the high water content of the material, the signal would likely pass directly through to the cladding without ever sensing the coating surface. X-ray fluorescence was also identified as a candidate technique, but would not work because the high gamma background from the irradiated fuel would swamp out the low energy aluminum signal. Laser interferometry could possibly be applied, but considerable development would be required and it was considered to be high risk on a short term basis. The consensus reached was that standard eddy current techniques for coating thickness measurement had the best chance for

  14. Transfer of elements relevant to nuclear fuel cycle from soil to boreal plants and animals in experimental meso- and microcosms.

    PubMed

    Tuovinen, Tiina S; Kasurinen, Anne; Häikiö, Elina; Tervahauta, Arja; Makkonen, Sari; Holopainen, Toini; Juutilainen, Jukka

    2016-01-01

    Uranium (U), cobalt (Co), molybdenum (Mo), nickel (Ni), lead (Pb), thorium (Th) and zinc (Zn) occur naturally in soil but their radioactive isotopes can also be released into the environment during the nuclear fuel cycle. The transfer of these elements was studied in three different trophic levels in experimental mesocosms containing downy birch (Betula pubescens), narrow buckler fern (Dryopteris carthusiana) and Scandinavian small-reed (Calamagrostis purpurea ssp. Phragmitoides) as producers, snails (Arianta arbostorum) as herbivores, and earthworms (Lumbricus terrestris) as decomposers. To determine more precisely whether the element uptake of snails is mainly via their food (birch leaves) or both via soil and food, a separate microcosm experiment was also performed. The element uptake of snails did not generally depend on the presence of soil, indicating that the main uptake route was food, except for U, where soil contact was important for uptake when soil U concentration was high. Transfer of elements from soil to plants was not linear, i.e. it was not correctly described by constant concentration ratios (CR) commonly applied in radioecological modeling. Similar nonlinear transfer was found for the invertebrate animals included in this study: elements other than U were taken up more efficiently when element concentration in soil or food was low. PMID:26363398

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

    DOEpatents

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

    1999-01-01

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

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

    DOEpatents

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

    1999-03-23

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

  17. Combined catalysts for the combustion of fuel in gas turbines

    DOEpatents

    Anoshkina, Elvira V.; Laster, Walter R.

    2012-11-13

    A catalytic oxidation module for a catalytic combustor of a gas turbine engine is provided. The catalytic oxidation module comprises a plurality of spaced apart catalytic elements for receiving a fuel-air mixture over a surface of the catalytic elements. The plurality of catalytic elements includes at least one primary catalytic element comprising a monometallic catalyst and secondary catalytic elements adjacent the primary catalytic element comprising a multi-component catalyst. Ignition of the monometallic catalyst of the primary catalytic element is effective to rapidly increase a temperature within the catalytic oxidation module to a degree sufficient to ignite the multi-component catalyst.

  18. Emission estimates of organic and elemental carbon from household biomass fuel used over the Indo-Gangetic Plain (IGP), India

    NASA Astrophysics Data System (ADS)

    Saud, T.; Gautam, R.; Mandal, T. K.; Gadi, Ranu; Singh, D. P.; Sharma, S. K.; Dahiya, Manisha; Saxena, M.

    2012-12-01

    Biomass burning emits large amount of aerosols and trace gases into the atmosphere, which have significant impact on atmospheric chemistry and climate. In the present study, we have selected seven Indian states (Delhi, Punjab, Haryana, Uttar Pradesh, Uttarakhand, Bihar and West Bengal) over the IGP, India. Samples of biomass fuel (Fuel Wood, Crop Residue and Dung Cake) from rural household have been collected (Saud et al., 2011a). The burning process has been simulated using a dilution sampler following the methodology developed by Venkatraman et al. (2005). In the present study, emission factor represents the total period of burning including pyrolysis, flaming and smoldering. We have determined the emission factors of organic carbon (OC) and elemental carbon (EC) from different types of biomass fuels collected over the study area. Average emission factors of OC from dung cake, fuel wood and crop residue over IGP, India are estimated as 3.87 ± 1.09 g kg-1, 0.95 ± 0.27 g kg-1, 1.46 ± 0.73 g kg-1, respectively. Similarly, average emission factors of EC from dung cake, fuel wood and crop residue over IGP, India are found to be 0.49 ± 0.25 g kg-1, 0.35 ± 0.07 g kg-1 and 0.37 ± 0.14 g kg-1, respectively. Dung cake and crop residue are normally not used in Uttarakhand. Annual budget of OC and EC from biomass fuels used as energy in rural households of IGP, India is estimated as 361.96 ± 170.18 Gg and 56.44 ± 29.06 Gg respectively. This study shows the regional emission inventory from Indian scenario with spatial variability.

  19. Criticality Safety of Low-Enriched Uranium and High-Enriched Uranium Fuel Elements in Heavy Water Lattices

    SciTech Connect

    Pesic, Milan P

    2003-10-15

    The RB reactor was designed as a natural-uranium, heavy water, nonreflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments', contributing to the preservation of a rather small number of heavy water benchmark critical experiments.

  20. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    NASA Technical Reports Server (NTRS)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  1. New fuel management plan for the Penn State TRIGA

    SciTech Connect

    Hughes, D.; Boyle, P.; Levine, S.H.

    1996-12-31

    The Pennsylvania State University (PSU) Breazeale TRIGA has utilized 12 wt% U fuel in the core since July 1992, when six 12 wt% U fuel elements were loaded to replace the depleted 8.5 wt% U fuel in the centermost ring, the B ring. This reload increased the cold k{sub eff} from 1.03 to 1.05, the cold k{sub eff} of 1.03 being the minimum k{sub eff} that will permit 1-MW operation for a sustained period. In the next fuel reload, this 12 wt% U fuel is to be moved outward to the adjacent ring, the C ring, and six fresh 12 wt% U fuel elements are to be added to the B ring. It was determined that using the 12 wt% U in place of 8.5 wt% U fuel reduced fuel costs by a factor of 6, and continuing this use of six 12 wt% U fuel elements for each reload maintained the lower fuel costs. This reloading technique worked successfully, requiring only 26 additional 12 wt% U elements to be loaded into the core during the last 23 yr. Recently, however, a new instrumented 12 wt% U fuel element read much higher temperatures than all previous similar fuel elements. Its measured fuel temperature at 1 MW is 585{degrees}C. As a result, the PSU TRIGA now operates at or below 60% full power to prevent this element from reaching fuel temperatures well above 500{degrees}C. The purpose of this paper is to describe a new fuel management strategy developed to use 12 wt% U fuel, which permits 1-MW operation and limits the maximum fuel temperature to {approx}500{degrees}C.

  2. FEM (finite element method) thermal modeling and thermal hydraulic performance of an enhanced thermal conductivity UO2/BeO composite fuel

    SciTech Connect

    Zhou, Wenzhong

    2011-03-24

    An enhanced thermal conductivity UO2-BeO composite nuclear fuel was studied. A methodology to generate ANSYS (an engineering simulation software) FEM (Finite Element Method) thermal models of enhanced thermal conductivity oxide nuclear fuels was developed. The results showed significant increase in the fuel thermal conductivities and have good agreement with the measured ones. The reactor performance analysis showed that the decrease in centerline temperature was 250-350K for the UO2-BeO composite fuel, and thus we can improve nuclear reactors' performance and safety, and high-level radioactive waste generation.

  3. FINITE ELEMENT SIMULATION FOR STRUCTURAL RESPONSE OF U7MO DISPERSION FUEL PLATES VIA FLUID-THERMAL-STRUCTURAL INTERACTION

    SciTech Connect

    Hakan Ozaltun; Herman Shen; Pavel Madvedev

    2010-11-01

    This article presents numerical simulation of dispersion fuel mini plates via fluid–thermal–structural interaction performed by commercial finite element solver COMSOL Multiphysics to identify initial mechanical response under actual operating conditions. Since fuel particles are dispersed in Aluminum matrix, and temperatures during the fabrication process reach to the melting temperature of the Aluminum matrix, stress/strain characteristics of the domain cannot be reproduced by using simplified models and assumptions. Therefore, fabrication induced stresses were considered and simulated via image based modeling techniques with the consideration of the high temperature material data. In order to identify the residuals over the U7Mo particles and the Aluminum matrix, a representative SEM image was employed to construct a microstructure based thermo-elasto-plastic FE model. Once residuals and plastic strains were identified in micro-scale, solution was used as initial condition for subsequent multiphysics simulations at the continuum level. Furthermore, since solid, thermal and fluid properties are temperature dependent and temperature field is a function of the velocity field of the coolant, coupled multiphysics simulations were considered. First, velocity and pressure fields of the coolant were computed via fluidstructural interaction. Computed solution for velocity fields were used to identify the temperature distribution on the coolant and on the fuel plate via fluid-thermal interaction. Finally, temperature fields and residual stresses were used to obtain the stress field of the plates via fluid-thermal-structural interaction.

  4. Molten tin reprocessing of spent nuclear fuel elements. [Patent application; continuous process

    DOEpatents

    Heckman, R.A.

    1980-12-19

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

  5. Computation of Dancoff Factors for Fuel Elements Incorporating Randomly Packed TRISO Particles

    SciTech Connect

    J. L. Kloosterman; Abderrafi M. Ougouag

    2005-01-01

    A new method for estimating the Dancoff factors in pebble beds has been developed and implemented within two computer codes. The first of these codes, INTRAPEB, is used to compute Dancoff factors for individual pebbles taking into account the random packing of TRISO particles within the fuel zone of the pebble and explicitly accounting for the finite geometry of the fuel kernels. The second code, PEBDAN, is used to compute the pebble-to-pebble contribution to the overall Dancoff factor. The latter code also accounts for the finite size of the reactor vessel and for the proximity of reflectors, as well as for fluctuations in the pebble packing density that naturally arises in pebble beds.

  6. Analysis of Topaz-II thermionic fuel element performance using TFEHX

    SciTech Connect

    Klein, A.C. ); Pawlowski, R.A. )

    1993-01-20

    Data reported by Russian Scientists and engineers for the TOPAZ-II single cell thermionic fuel elments (TFE) is compared with analytical results calculated using the TFEHX computer program in order to benchmark the code. The results of this comparison show good agreement with the TOPAZ-II results over a wide range of power inputs, cesium vapor pressures, and other design variables. Future refinements of the TFEHX methodology should enhance the performance of the code to better predict single cell TFE behavior.

  7. The use of U/sub 3/Si/sub 2/ dispersed in aluminum in plate-type fuel elements for research and test reactors

    SciTech Connect

    Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

    1987-10-01

    A high-density fuel based on U/sub 3/Si/sub 2/ dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U/sub 3/Si/sub 2/ fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U/sub 3/Si/sub 2/ particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U/sub 3/Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U/sub 3/Si/sub 2/-aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m/sup 3/ is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs.

  8. Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

    SciTech Connect

    Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A; Drypolcher, Anthony F; Hickey, Joseph

    2010-01-01

    We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is in support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to consolidate the

  9. A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

    SciTech Connect

    Senor, David J.; Painter, Chad L.; Geelhood, Ken J.; Wootan, David W.; Meriwether, George H.; Cuta, Judith M.; Adkins, Harold E.; Matson, Dean W.; Abrego, Celestino P.

    2007-12-01

    Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling, core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.

  10. Preparation, loading and storage of castor THTR/AVR-casks for spent fuel elements-part of the decommissioning of the high temperature reactor AVR

    SciTech Connect

    Theenhaus, R.; Halaszovich, S.; Storch, S.

    1994-12-31

    Shipment of spent fuel elements of the AVR reactor to the interim storage site at the Forschungszentrum Juelich KFA has started in August 1993. The handling procedure involves the preparation of the transport-storage casks, their remote loading with two stainless steel flasks, each of them containing 950 spherical fuel elements, the closing of the casks, the leak testing, the dose rate measurements, smear tests, the transportation and stacking of the casks and finally their connection with the permanent electronical leak control system. The handling strictly follows the manual which is part of the license and it is inspected by a member of the supervising ministry and an independent expert ordered by the ministry. Until the end of 1993 32,300 fuel elements have been transferred to the storage site.

  11. Discrete Element Model for Simulations of Early-Life Thermal Fracturing Behaviors in Ceramic Nuclear Fuel Pellets

    SciTech Connect

    Hai Huang; Ben Spencer; Jason Hales

    2014-10-01

    A discrete element Model (DEM) representation of coupled solid mechanics/fracturing and heat conduction processes has been developed and applied to explicitly simulate the random initiations and subsequent propagations of interacting thermal cracks in a ceramic nuclear fuel pellet during initial rise to power and during power cycles. The DEM model clearly predicts realistic early-life crack patterns including both radial cracks and circumferential cracks. Simulation results clearly demonstrate the formation of radial cracks during the initial power rise, and formation of circumferential cracks as the power is ramped down. In these simulations, additional early-life power cycles do not lead to the formation of new thermal cracks. They do, however clearly indicate changes in the apertures of thermal cracks during later power cycles due to thermal expansion and shrinkage. The number of radial cracks increases with increasing power, which is consistent with the experimental observations.

  12. ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

    NASA Technical Reports Server (NTRS)

    Walton, J. T.

    1994-01-01

    ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

  13. Numerical analysis of a nuclear fuel element for nuclear thermal propulsion

    NASA Technical Reports Server (NTRS)

    Wang, Ten-See; Schutzenhofer, Luke

    1991-01-01

    A computational fluid dynamics model with porosity and permeability formulations in the transport equations has been developed to study the concept of nuclear thermal propulsion through the analysis of a pulsed irradiation of a particle bed element (PIPE). The numerical model is a time-accurate pressure-based formulation. An adaptive upwind scheme is employed for spatial discretization. The upwind scheme is based on second- and fourth-order central differencing with adaptive artificial dissipation. Multiblocked porosity regions have been formulated to model the cold frit, particle bed, and hot frit. Multiblocked permeability regions have been formulated to describe the flow shaping effect from the thickness-varying cold frit. Computational results for several zero-power density PIPEs and an elevated-particle-temperature PIPE are presented. The implications of the computational results are discussed.

  14. Selective Catalytic Oxidation of Hydrogen Sulfide to Elemental Sulfur in the Presence of Coal-Derived Fuel Gas

    SciTech Connect

    Stevens, R.W., Jr.; Gardner, T.H.; Shekhawat, Dushyant; Berry, D.A.; Freed, A.D.

    2005-09-01

    A rotatable central composite design approach was utilized to examine the activity and selectivity of an activated carbon-based catalyst during partial oxidation of H2S to elemental sulfur. Tests were conducted at 400 psig with a 4000 ppmv H2S inlet concentration in the presence of coal-derived fuel gas (i.e., CO, CO2, H2, and H2O). Temperature, space velocity, and O2:H2S ratio were chosen as process variables for the study to generate response surface maps of elemental sulfur yield and longevity. Temperature was varied from 150-300°C; gas hourly space velocity varied from 1000 -15000 h-1 (STP); O2:H2S varied from 0.5-2.0. Evolution of SO2 and/or COS during some of the tests indicated the presence of side reactions. The relationship between the process variables and catalyst performance as well as the results of catalyst characterization is discussed.

  15. Trace element partitioning in ashes from boilers firing pure wood or mixtures of solid waste with respect to fuel composition, chlorine content and temperature.

    PubMed

    Saqib, Naeem; Bäckström, Mattias

    2014-12-01

    Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of fly ash and bottom ash were collected from six incineration facilities across Sweden including two grate fired and four fluidized bed incinerators, to have a variation in the input fuel composition (from pure biofuel to mixture of waste) and different temperature boiler conditions. As trace element concentrations in the input waste at the same facilities have already been analyzed, the present study focuses on the concentration of trace elements in the waste fuel, their distribution in the incineration residues with respect to chlorine content of waste and combustion temperature. Results indicate that Zn, Cu and Pb are dominating trace elements in the waste fuel. Highly volatile elements mercury and cadmium are mainly found in fly ash in all cases; 2/3 of lead also end up in fly ash while Zn, As and Sb show a large variation in distribution with most of them residing in the fly ash. Lithophilic elements such as copper and chromium are mainly found in bottom ash from grate fired facilities while partition mostly into fly ash from fluidized bed incinerators, especially for plants fuelled by waste wood or ordinary wood chips. There is no specific correlation between input concentration of an element in the waste fuel and fraction partitioned to fly ash. Temperature and chlorine content have significant effects on partitioning characteristics by increasing the formation and vaporization of highly volatile metal chlorides. Zinc and cadmium concentrations in fly ash increase with the incineration temperature. PMID:25263218

  16. Molten carbonate fuel cell separator

    DOEpatents

    Nickols, Richard C.

    1986-09-02

    In a stacked array of molten carbonate fuel cells, a fuel cell separator is positioned between adjacent fuel cells to provide isolation as well as a conductive path therebetween. The center portion of the fuel cell separator includes a generally rectangular, flat, electrical conductor. Around the periphery of the flat portion of the separator are positioned a plurality of elongated resilient flanges which form a gas-tight seal around the edges of the fuel cell. With one elongated flange resiliently engaging a respective edge of the center portion of the separator, the sealing flanges, which are preferably comprised of a noncorrosive material such as an alloy of yttrium, iron, aluminum or chromium, form a tight-fitting wet seal for confining the corrosive elements of the fuel cell therein. This arrangement permits a good conductive material which may be highly subject to corrosion and dissolution to be used in combination with a corrosion-resistant material in the fuel cell separator of a molten carbonate fuel cell for improved fuel cell conductivity and a gas-tight wet seal.

  17. Molten carbonate fuel cell separator

    DOEpatents

    Nickols, R.C.

    1984-10-17

    In a stacked array of molten carbonate fuel cells, a fuel cell separator is positioned between adjacent fuel cells to provide isolation as well as a conductive path therebetween. The center portion of the fuel cell separator includes a generally rectangular, flat, electrical conductor. Around the periphery of the flat portion of the separator are positioned a plurality of elongated resilient flanges which form a gas-tight seal around the edges of the fuel cell. With one elongated flange resiliently engaging a respective edge of the center portion of the separator, the sealing flanges, which are preferably comprised of a noncorrosive material such as an alloy of yttrium, iron, aluminum or chromium, form a tight-fitting wet seal for confining the corrosive elements of the fuel cell therein. This arrangement permits a good conductive material which may be highly subject to corrosion and dissolution to be used in combination with a corrosion-resistant material in the fuel cell separator of a molten carbonate fuel cell for improved fuel cell conductivity and a gas-tight wet seal.

  18. Simulated Verification of Fuel Element Inventory in a Small Reactor Core Using the Nuclear Materials Identification System (NMIS)

    SciTech Connect

    Grogan, Brandon R; Mihalczo, John T

    2009-01-01

    The International Panel on Climate Change projects that by 2050 the world energy demand may double. Although the primary focus for new nuclear power plants in industrialized nations is on large plants in the 1000-1600 MWe range, there is an increasing demand for small and medium reactors (SMRs). About half of the innovative SMR concepts are for small (<300 MWe) reactors with a 5-30 year life without on-site refueling. This type of reactor is also known as a battery-type reactor. These reactors are particularly attractive to countries with small power grids and for non-electrical purposes such as heating, hydrogen production, and seawater desalination. Traditionally, this type of reactor has been used in a nautical propulsion role. This type of reactor is designed as a permanently sealed unit to prevent the diversion of the uranium in the core by the user. However, after initial fabrication it will be necessary to verify that the newly fabricated reactor core contains the quantity of uranium that initially entered the fuel fabrication plant. In most instances, traditional inspection techniques can be used to perform this verification, but in certain situations the core design will be considered sensitive. Non-intrusive verification techniques must be utilized in these situations. The Nuclear Materials Identification System (NMIS) with imaging uses active interrogation and a fast time correlation processor to characterize fissile material. The MCNP-PoliMi computer code was used to simulate NMIS measurements of a small, sealed reactor core. Because most battery-type reactor designs are still in the early design phase, a more traditional design based on a Russian icebreaker core was used in the simulations. These simulations show how the radiography capabilities of the NMIS could be used to detect the diversion of fissile material by detecting void areas in the assembled core where fuel elements have been removed.

  19. Metallic elements in fossil fuel combustion products: amounts and form of emissions and evaluation of carcinogenicity and mutagenicity.

    PubMed Central

    Vouk, V B; Piver, W T

    1983-01-01

    Metallic elements contained in coal, oil and gasoline are mobilized by combustion processes and may be emitted into the atmosphere, mainly as components of submicron particles. The information about the amounts, composition and form of metal compounds is reviewed for some fuels and combustion processes. Since metal compounds are always contained in urban air pollutants, they have to be considered whenever an evaluation of biological impact of air pollutants is made. The value of currently used bioassays for the evaluation of the role of trace metal compounds, either as major biologically active components or as modifiers of biological effects of organic compounds is assessed. The whole animal bioassays for carcinogenicity do not seem to be an appropriate approach. They are costly, time-consuming and not easily amenable to the testing of complex mixtures. Some problems related to the application and interpretation of short-term bioassays are considered, and the usefulness of such bioassays for the evaluation of trace metal components contained in complex air pollution mixtures is examined. PMID:6337825

  20. A combined Cyanex-923/HEH[EHP]/Dodecane solvent for recovery of transuranic elements from used nuclear fuel

    SciTech Connect

    Johnson, A.; Nash, K.L.

    2013-07-01

    The separation of minor actinides from fission product lanthanides remains a primary challenge for enabling the recycle of used nuclear fuel. To minimize the complexity of materials handling, combining extractant processes has become an increasingly attractive option. Unfortunately, combined processes sometimes suffer reduced utility due to strong dipole-dipole interactions between the extractants. The results reported here describe a system based on a combination of commercially available extractants Cyanex-923 and HEH[EHP]. In contrast to other combined extractant systems, these extractant molecules exhibit comparatively weak interactions, reducing the impact of secondary interactions. In this process, mixtures containing equal ratios of Cyanex-923 and HEH[EHP] were seen to co-extract americium and the lanthanides from nitric acid solutions. Stripping of An(III) was effectively achieved through contact with an aqueous phase comprised of glycine (for pH control) and a polyamino-poly-carboxylate stripping reagent that selectively removes An(III) from the extractant phase. The lanthanides can then be stripped from the loaded organic phase contacting with high nitric acid concentrations. Extraction of fission products zirconium and molybdenum was also investigated and potential strategies for their management have been identified. The work presented demonstrates the feasibility of combining Cyanex-923 and HEH[EHP] for separating and recovering the transuranic elements from the Ln(III). (authors)

  1. Comparison between steady-state and dynamic I-V measurements from a single-cell thermionic fuel element

    SciTech Connect

    Wernsman, B.

    1997-01-01

    A comparison between steady-state and dynamic I-V measurements from a single-cell thermionic fuel element (TFE) is made. The single-cell TFE used in this study is the prototype for the 40kW{sub e} space nuclear power system that is similar to the 6kW{sub e} TOPAZ-II. The steady-state I-V measurements influence the emitter temperature due to electron cooling. Therefore, to eliminate the steady-state I-V measurement influence on the TFE and provide a better understanding of the behavior of the thermionic energy converter and TFE characteristics, dynamic I-V measurements are made. The dynamic I-V measurements are made at various input power levels, cesium pressures, collector temperatures, and steady-state current levels. From these measurements, it is shown that the dynamic I-V{close_quote}s do not change the TFE characteristics at a given operating point. Also, the evaluation of the collector work function from the dynamic I-V measurements shows that the collector optimization is not due to a minimum in the collector work function but due to an emission optimization. Since the dynamic I-V measurements do not influence the TFE characteristics, it is believed that these measurements can be done at a system level to understand the influence of TFE placement in the reactor as a function of the core thermal distribution. {copyright} {ital 1997 American Institute of Physics.}

  2. Comparison between steady-state and dynamic I-V measurements from a single-cell thermionic fuel element

    SciTech Connect

    Wernsman, Bernard

    1997-01-10

    A comparison between steady-state and dynamic I-V measurements from a single-cell thermionic fuel element (TFE) is made. The single-cell TFE used in this study is the prototype for the 40 kW{sub e} space nuclear power system that is similar to the 6 kW{sub e} TOPAZ-II. The steady-state I-V measurements influence the emitter temperature due to electron cooling. Therefore, to eliminate the steady-state I-V measurement influence on the TFE and provide a better understanding of the behavior of the thermionic energy converter and TFE characteristics, dynamic I-V measurements are made. The dynamic I-V measurements are made at various input power levels, cesium pressures, collector temperatures, and steady-state current levels. From these measurements, it is shown that the dynamic I-V's do not change the TFE characteristics at a given operating point. Also, the evaluation of the collector work function from the dynamic I-V measurements shows that the collector optimization is not due to a minimum in the collector work function but due to an emission optimization. Since the dynamic I-V measurements do not influence the TFE characteristics, it is believed that these measurements can be done at a system level to understand the influence of TFE placement in the reactor as a function of the core thermal distribution.

  3. Features of temperature control of fuel element cladding for pressurized water nuclear reactor ``WWER-1000'' while simulating reactor accidents

    NASA Astrophysics Data System (ADS)

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M.

    2013-09-01

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones "cladding-junction" and "junction-coolant". The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ˜5% of maximum value by the final moment of the stage of linear increase in the temperature.

  4. Determination of uranium in urine samples of fuel element fabrication workers by beta-delayed neutron counting

    NASA Astrophysics Data System (ADS)

    Gabelmann, H.; Lerch, M.; Kratz, K.-L.; Rudolph, W.

    1984-06-01

    Within the health physics examination of fuel element fabrication workers, the control of uranium incorporation is of importance. This is commonly performed by the determination of the alpha activity concentration of uranium excreted in the urine. However, since the chemical separation procedure and the preparation of alpha-counting samples are complicated and time-consuming, this method may imply restrictions on the routine control of large numbers of persons. Therefore, we have investigated the applicability of measuring the beta-delayed neutrons from thermal neutron induced fission of the 235U in the urine samples. The uranium was separated by coprecipitation with Fe(OH) 3 from the urine samples and irradiated in a rabbit system of the Mainz TRIGA reactor. The neutrons were counted with a 3He long counter. The detection limit of 0.3 to 0.9 pCi 1 -1 is comparable to that of alpha spectrometry, but the time required for one sample, from preparation to data evaluation is less than 25 min.

  5. 46 CFR 148.445 - Adjacent spaces.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 5 2013-10-01 2013-10-01 false Adjacent spaces. 148.445 Section 148.445 Shipping COAST... THAT REQUIRE SPECIAL HANDLING Additional Special Requirements § 148.445 Adjacent spaces. When... following requirements must be met: (a) Each space adjacent to a cargo hold must be ventilated by...

  6. 46 CFR 148.445 - Adjacent spaces.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 5 2014-10-01 2014-10-01 false Adjacent spaces. 148.445 Section 148.445 Shipping COAST... THAT REQUIRE SPECIAL HANDLING Additional Special Requirements § 148.445 Adjacent spaces. When... following requirements must be met: (a) Each space adjacent to a cargo hold must be ventilated by...

  7. 46 CFR 148.445 - Adjacent spaces.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 5 2012-10-01 2012-10-01 false Adjacent spaces. 148.445 Section 148.445 Shipping COAST... THAT REQUIRE SPECIAL HANDLING Additional Special Requirements § 148.445 Adjacent spaces. When... following requirements must be met: (a) Each space adjacent to a cargo hold must be ventilated by...

  8. 46 CFR 148.445 - Adjacent spaces.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 5 2011-10-01 2011-10-01 false Adjacent spaces. 148.445 Section 148.445 Shipping COAST... THAT REQUIRE SPECIAL HANDLING Additional Special Requirements § 148.445 Adjacent spaces. When... following requirements must be met: (a) Each space adjacent to a cargo hold must be ventilated by...

  9. Comparative finite element analysis of the stress-strain states in three different bonded solid oxide fuel cell seal designs

    NASA Astrophysics Data System (ADS)

    Weil, K. S.; Koeppel, B. J.

    One of the critical issues in designing and fabricating a high performance planar solid oxide fuel cell (pSOFC) stack is the development of the appropriate materials and techniques for hermetically sealing the metal and ceramic components. A second critical issue is ensuring that the brittle ceramic cell constituents, i.e. the electrodes and electrolyte, exhibit high mechanical reliability by mitigating potential sources of thermal-mechanically induced stresses that can lead to fracture during operation and/or shutdown. A foil-based sealing approach is currently being developed that appears to offer good hermeticity and mechanical integrity, while minimizing the generation of high stresses in either of the joint's substrate materials. Based on the concept's viability, demonstrated in prior experimental work, numerical analyses were conducted to evaluate the behavior and benefits of the seal in a configuration prototypic of current pSOFC stack designs. This paper presents recent results from finite element (FE) simulations of a planar cell using the foil-based seal, along with companion analyses of the more conventionally employed glass-ceramic and brazed joints. The stresses and deformations of the components were evaluated at isothermal operating and shutdown temperatures. The results indicate that the foil seal is able to accommodate a significant degree of thermal mismatch strain between the metallic support structure and the ceramic cell via elastic deformations of the foil and plasticity in the foil-to-cell braze layer. Consequently the cell stresses in this type of seal are predicted to be much lower than those in the glass-ceramic and brazed designs, which is expected to lead to improved stack reliability. This ability to accommodate large thermal strain mismatches allows the design requirement of thermal expansion matching between ceramic and metal stack components to be relaxed and expands the list of candidate materials that can be considered for the

  10. Trace element partitioning in ashes from boilers firing pure wood or mixtures of solid waste with respect to fuel composition, chlorine content and temperature

    SciTech Connect

    Saqib, Naeem Bäckström, Mattias

    2014-12-15

    Highlights: • Different solids waste incineration is discussed in grate fired and fluidized bed boilers. • We explained waste composition, temperature and chlorine effects on metal partitioning. • Excessive chlorine content can change oxide to chloride equilibrium partitioning the trace elements in fly ash. • Volatility increases with temperature due to increase in vapor pressure of metals and compounds. • In Fluidized bed boiler, most metals find themselves in fly ash, especially for wood incineration. - Abstract: Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of fly ash and bottom ash were collected from six incineration facilities across Sweden including two grate fired and four fluidized bed incinerators, to have a variation in the input fuel composition (from pure biofuel to mixture of waste) and different temperature boiler conditions. As trace element concentrations in the input waste at the same facilities have already been analyzed, the present study focuses on the concentration of trace elements in the waste fuel, their distribution in the incineration residues with respect to chlorine content of waste and combustion temperature. Results indicate that Zn, Cu and Pb are dominating trace elements in the waste fuel. Highly volatile elements mercury and cadmium are mainly found in fly ash in all cases; 2/3 of lead also end up in fly ash while Zn, As and Sb show a large variation in distribution with most of them residing in the fly ash. Lithophilic elements such as copper and chromium are mainly found in bottom ash from grate fired facilities while partition mostly into fly ash from fluidized bed incinerators, especially for plants fuelled by waste wood or ordinary wood chips. There is no specific correlation between input concentration of an element in the waste fuel and fraction partitioned to fly ash. Temperature and chlorine

  11. The Manufacture of W-UO2 Fuel Elements for NTP Using the Hot Isostatic Pressing Consolidation Process

    NASA Technical Reports Server (NTRS)

    Broadway, Jeramie; Hickman, Robert; Mireles, Omar

    2012-01-01

    NTP is attractive for space exploration because: (1) Higher Isp than traditional chemical rockets (2)Shorter trip times (3) Reduced propellant mass (4) Increased payload. Lack of qualified fuel material is a key risk (cost, schedule, and performance). Development of stable fuel form is a critical path, long lead activity. Goals of this project are: Mature CERMET and Graphite based fuel materials and Develop and demonstrate critical technologies and capabilities.

  12. The development of a method for the determination of trace elements in fuel alcohol by electrothermal vaporization inductively coupled plasma mass spectrometry using external calibration

    NASA Astrophysics Data System (ADS)

    Saint'Pierre, Tatiana Dillenburg; Maranhão, Tatiane de Andrade; Frescura, Vera Lúcia Azzolin; Curtius, Adilson José

    2005-06-01

    A method for the determination of Ag, As, Cd, Cu, Co, Fe, Mn, Ni, Pb, Sn and Tl in fuel alcohol by electrothermal vaporization inductively coupled plasma mass spectrometry is proposed. The determinations were carried out by external calibration against ethanolic solutions, without a chemical modifier, employing the following pyrolysis and vaporization temperatures: 400 °C and 2300 °C for the more volatile analytes and 1000 °C and 2500 °C for the less volatile analytes. The determination of As, Cd, Pb, Sn and Tl was additionally carried out using Pd as modifier at 800 °C pyrolysis and 2400 °C vaporization temperatures. The temperatures were optimized through pyrolysis and vaporization curves. Seven common fuel ethanol, one fuel ethanol with additive and one anhydrous fuel ethanol sample have been analyzed. The measured concentrations were at the μg L -1 level or lower. Since there is no certified reference material for fuel ethanol, the accuracy of the method was checked by the recovery test, with recoveries from 75% to 124%. The limits of detection (LODs), in μg L -1, and the relative standard deviations for 5 replicates were, for the elements in the conditions without modifier: Ag: 0.015 and 9.1%, Co: 0.002 and 10%, Cu: 0.22 and 6.6%, Fe: 0.72 and 4.3%, Mn: 0.025 and 12%, Ni: 0.026 and 9.3%, and for the elements with Pd: As: 0.02 and 2.9%, Cd: 0.07 and 25%, Pb: 0.02 and 3.1%, Sn: 0.010 and 6.0%, Tl: 0.0008 and 2.5%. Electrothermal vaporization avoids the loading of the plasma with organics, allowing the analysis of fuel ethanol by ICP-MS with good accuracy and reasonable precision.

  13. A prototype expert system for the monitoring of defected nuclear fuel elements in Canada deuterium uranium reactors

    SciTech Connect

    Lewis, B.J.; Green, R.J. ); Che, C.W.T. )

    1992-06-01

    This paper reports on a prototype expert system for fuel failure monitoring in Canada deuterium uranium (CANDU) power reactors. Based on a coolant activity analysis, the system is able to provide information in an operating reactor on the number of fuel failures, the average defect size, and the amount of tramp uranium deposited on the in-core surfaces of the primary heat transport system. The fission product release model used in the system is based on results from an in-reactor experimental program at Chalk River Nuclear Laboratories. The expert system is validated against fuel failure data from a number of CANDU power reactors.

  14. Combining octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide and bis-(2-ethylhexyl)phosphoric acid extractants for recovering transuranic elements from irradiated nuclear fuel

    SciTech Connect

    Lumetta, Gregg J.; Carter, Jennifer C.; Gelis, Artem V.; Vandegrift, George F.

    2009-10-14

    Advanced concepts for closing the nuclear fuel cycle include separating Am and Cm from other fuel components. Separating these elements from the lanthanide elements at an industrial scale remains a significant technical challenge. We describe here a chemical system in which a neutral extractant--octyl(phenyl)-N,N-diisobutyl-carbamoylmethyl-phosphine oxide (CMPO)--is combined with an acidic extractant--bis-(2-ethylhexyl)phosphoric acid (HDEHP)--to form a single process solvent (with dodecane as the diluent) for separating Am and Cm from the other components of irradiated nuclear fuel. Continuous variation experiments in which the relative CMPO and HDEHP concentrations are varied indicate a synergistic relationship between the two extractants in the extraction of Am from buffered diethylenetriaminepentaacetic acid (DTPA) solutions. A solvent mixture consisting or 0.1 M CMPO + 1 M HDEHP in dodecane offers acceptable extraction efficiency for the trivalent lanthanides and actinides from 1 M HNO3 while maintaining good lanthanide/actinide separation factors in the stripping regime (buffered DTPA solutions with pH 3.5 to 4). Using citrate buffer instead of lactate buffer results in improved lanthanide/actinide separation factors.

  15. 43 CFR 420.3 - Adjacent lands.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 43 Public Lands: Interior 1 2014-10-01 2014-10-01 false Adjacent lands. 420.3 Section 420.3 Public Lands: Interior Regulations Relating to Public Lands BUREAU OF RECLAMATION, DEPARTMENT OF THE INTERIOR...-managing agencies on adjacent lands (both public and private)....

  16. 43 CFR 420.3 - Adjacent lands.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 43 Public Lands: Interior 1 2013-10-01 2013-10-01 false Adjacent lands. 420.3 Section 420.3 Public Lands: Interior Regulations Relating to Public Lands BUREAU OF RECLAMATION, DEPARTMENT OF THE INTERIOR...-managing agencies on adjacent lands (both public and private)....

  17. 43 CFR 420.3 - Adjacent lands.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 43 Public Lands: Interior 1 2011-10-01 2011-10-01 false Adjacent lands. 420.3 Section 420.3 Public Lands: Interior Regulations Relating to Public Lands BUREAU OF RECLAMATION, DEPARTMENT OF THE INTERIOR OFF-ROAD VEHICLE USE § 420.3 Adjacent lands. When administratively feasible, the regulation of off-road vehicle use on Reclamation lands will...

  18. 43 CFR 420.3 - Adjacent lands.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 43 Public Lands: Interior 1 2012-10-01 2011-10-01 true Adjacent lands. 420.3 Section 420.3 Public Lands: Interior Regulations Relating to Public Lands BUREAU OF RECLAMATION, DEPARTMENT OF THE INTERIOR...-managing agencies on adjacent lands (both public and private)....

  19. 43 CFR 420.3 - Adjacent lands.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 43 Public Lands: Interior 1 2010-10-01 2010-10-01 false Adjacent lands. 420.3 Section 420.3 Public Lands: Interior Regulations Relating to Public Lands BUREAU OF RECLAMATION, DEPARTMENT OF THE INTERIOR OFF-ROAD VEHICLE USE § 420.3 Adjacent lands. When administratively feasible, the regulation of...

  20. Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

    SciTech Connect

    McCulloch, R.W.; Post, D.W.; Lovell, R.T.; Snyder, S.D.

    1981-04-01

    Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators.

  1. Molecular disorganization of axons adjacent to human lacunar infarcts

    PubMed Central

    Lee, Monica D.; Tung, Spencer; Vinters, Harry V.; Carmichael, S. Thomas

    2015-01-01

    Cerebral microvascular disease predominantly affects brain white matter and deep grey matter, resulting in ischaemic damage that ranges from lacunar infarcts to white matter hyperintensities seen on magnetic resonance imaging. These lesions are common and result in both clinical stroke syndromes and accumulate over time, resulting in cognitive deficits and dementia. Magnetic resonance imaging studies suggest that these lesions progress over time, accumulate adjacent to prior lesions and have a penumbral region susceptible to further injury. The pathological correlates of this adjacent injury in surviving myelinated axons have not been previously defined. In this study, we sought to determine the molecular organization of axons in tissue adjacent to lacunar infarcts and in the regions surrounding microinfarcts, by determining critical elements in axonal function: the morphology and length of node of Ranvier segments and adjacent paranodal segments. We examined post-mortem brain tissue from six patients with lacunar infarcts and tissue from two patients with autosomal dominant retinal vasculopathy and cerebral leukoencephalopathy (previously known as hereditary endotheliopathy with retinopathy, nephropathy and stroke) who accumulate progressive white matter ischaemic lesions in the form of lacunar and microinfarcts. In axons adjacent to lacunar infarcts yet extending up to 150% of the infarct diameter away, both nodal and paranodal length increase by ∼20% and 80%, respectively, reflecting a loss of normal cell-cell adhesion and signalling between axons and oligodendrocytes. Using premorbid magnetic resonance images, brain regions from patients with retinal vasculopathy and cerebral leukoencephalopathy that harboured periventricular white matter hyperintensities were selected and the molecular organization of axons was determined within these regions. As in regions adjacent to lacunar infarcts, nodal and paranodal length in white matter of these patients is

  2. COMPARISON OF PARTICLE SIZE DISTRIBUTIONS AND ELEMENTAL PARTITIONING FROM THE COMBUSTION OF PULVERIZED COAL AND RESIDUAL FUEL OIL

    EPA Science Inventory

    The paper gives results of experimental efforts in which three coals and a residual fuel oil were combusted in three different systems simulating process and utility boilers. Particloe size distributions (PSDs) were determined using atmospheric and low-pressure impaction, electr...

  3. A finite element analysis modeling tool for solid oxide fuel cell development: coupled electrochemistry, thermal and flow analysis in MARC ®

    NASA Astrophysics Data System (ADS)

    Khaleel, M. A.; Lin, Z.; Singh, P.; Surdoval, W.; Collin, D.

    A 3D simulation tool for modeling solid oxide fuel cells is described. The tool combines the versatility and efficiency of a commercial finite element analysis code, MARC ®, with an in-house developed robust and flexible electrochemical (EC) module. Based upon characteristic parameters obtained experimentally and assigned by the user, the EC module calculates the current density distribution, heat generation, and fuel and oxidant species concentration, taking the temperature profile provided by MARC ® and operating conditions such as the fuel and oxidant flow rate and the total stack output voltage or current as the input. MARC ® performs flow and thermal analyses based on the initial and boundary thermal and flow conditions and the heat generation calculated by the EC module. The main coupling between MARC ® and EC is for MARC ® to supply the temperature field to EC and for EC to give the heat generation profile to MARC ®. The loosely coupled, iterative scheme is advantageous in terms of memory requirement, numerical stability and computational efficiency. The coupling is iterated to self-consistency for a steady-state solution. Sample results for steady states as well as the startup process for stacks with different flow designs are presented to illustrate the modeling capability and numerical performance characteristic of the simulation tool.

  4. Proposed fuel cycle for the Integral Fast Reactor

    SciTech Connect

    Burris, L.; Walters, L.C.

    1985-01-01

    One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor.

  5. FINITE-ELEMENT ANALYSIS OF ROCK FALL ON UNCANISTERED FUEL WASTE PACKAGE DESIGNS (SCPB: N/A)

    SciTech Connect

    Z. Ceylan

    1996-10-18

    The objective of this analysis is to explore the Uncanistered Fuel (UCF) Tube Design waste package (WP) resistance to rock falls. This analysis will also be used to determine the size of rock that can strike the WP without causing failure in the containment barriers from a height based on the starter tunnel dimensions. The purpose of this analysis is to document the models and methods used in the calculations.

  6. Axially staggered seed-blanket reactor-fuel-module construction. [LWBR

    DOEpatents

    Cowell, G.K.; DiGuiseppe, C.P.

    1982-10-28

    A heterogeneous nuclear reactor of the seed-blanket type is provided wherein the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements.

  7. Discrete Element Modeling

    SciTech Connect

    Morris, J; Johnson, S

    2007-12-03

    The Distinct Element Method (also frequently referred to as the Discrete Element Method) (DEM) is a Lagrangian numerical technique where the computational domain consists of discrete solid elements which interact via compliant contacts. This can be contrasted with Finite Element Methods where the computational domain is assumed to represent a continuum (although many modern implementations of the FEM can accommodate some Distinct Element capabilities). Often the terms Discrete Element Method and Distinct Element Method are used interchangeably in the literature, although Cundall and Hart (1992) suggested that Discrete Element Methods should be a more inclusive term covering Distinct Element Methods, Displacement Discontinuity Analysis and Modal Methods. In this work, DEM specifically refers to the Distinct Element Method, where the discrete elements interact via compliant contacts, in contrast with Displacement Discontinuity Analysis where the contacts are rigid and all compliance is taken up by the adjacent intact material.

  8. On the time-course of adjacent and non-adjacent transposed-letter priming

    PubMed Central

    Ktori, Maria; Kingma, Brechtsje; Hannagan, Thomas; Holcomb, Phillip J.; Grainger, Jonathan

    2014-01-01

    We compared effects of adjacent (e.g., atricle-ARTICLE) and non-adjacent (e.g., actirle-ARTICLE) transposed-letter (TL) primes in an ERP study using the sandwich priming technique. TL priming was measured relative to the standard double-substitution condition. We found significantly stronger priming effects for adjacent transpositions than non-adjacent transpositions (with 2 intervening letters) in behavioral responses (lexical decision latencies), and the adjacent priming effects emerged earlier in the ERP signal, at around 200 ms post-target onset. Non-adjacent priming effects emerged about 50 ms later and were short-lived, being significant only in the 250-300 ms time-window. Adjacent transpositions on the other hand continued to produce priming in the N400 time-window (300-500 ms post-target onset). This qualitatively different pattern of priming effects for adjacent and non-adjacent transpositions is discussed in the light of different accounts of letter transposition effects, and the utility of drawing a distinction between positional flexibility and positional noise. PMID:25364497

  9. Year-round Source Contributions of Fossil Fuel and Biomass Combustion to Elemental Carbon on the North Slope Alaska Utilizing Radiocarbon Analysis

    NASA Astrophysics Data System (ADS)

    Barrett, T. E.; Gustafsson, O.; Winiger, P.; Moffett, C.; Back, J.; Sheesley, R. J.

    2015-12-01

    It is well documented that the Arctic has undergone rapid warming at an alarming rate over the past century. Black carbon (BC) affects the radiative balance of the Arctic directly and indirectly through the absorption of incoming solar radiation and by providing a source of cloud and ice condensation nuclei. Among atmospheric aerosols, BC is the most efficient absorber of light in the visible spectrum. The solar absorbing efficiency of BC is amplified when it is internally mixed with sulfates. Furthermore, BC plumes that are fossil fuel dominated have been shown to be approximately 100% more efficient warming agents than biomass burning dominated plumes. The renewal of offshore oil and gas exploration in the Arctic, specifically in the Chukchi Sea, will introduce new BC sources to the region. This study focuses on the quantification of fossil fuel and biomass combustion sources to atmospheric elemental carbon (EC) during a year-long sampling campaign in the North Slope Alaska. Samples were collected at the Department of Energy Atmospheric Radiation Measurement (ARM) climate research facility in Barrow, AK, USA. Particulate matter (PM10) samples collected from July 2012 to June 2013 were analyzed for EC and sulfate concentrations combined with radiocarbon (14C) analysis of the EC fraction. Radiocarbon analysis distinguishes fossil fuel and biomass burning contributions based on large differences in end members between fossil and contemporary carbon. To perform isotope analysis on EC, it must be separated from the organic carbon fraction of the sample. Separation was achieved by trapping evolved CO2 produced during EC combustion in a cryo-trap utilizing liquid nitrogen. Radiocarbon results show an average fossil contribution of 85% to atmospheric EC, with individual samples ranging from 47% to 95%. Source apportionment results will be combined with back trajectory (BT) analysis to assess geographic source region impacts on the EC burden in the western Arctic.

  10. SOLIDS TRANSPORT BETWEEN ADJACENT CAFB FLUIDIZED BEDS

    EPA Science Inventory

    The report gives results of an experimental investigation of a pulsed, dense-phase pneumatic transport system for controlled circulation between adjacent fluidized beds. A model was developed to predict performance. The program provides technical support for EPA's program to demo...

  11. Border separation for adjacent orthogonal fields

    SciTech Connect

    Werner, B.L.; Khan, F.M.; Sharma, S.C.; Lee, C.K.; Kim, T.H. )

    1991-06-01

    Field border separations for adjacent orthogonal fields can be calculated geometrically, given the validity of some important assumptions such as beam alignment and field uniformity. Thermoluminescent dosimetry (TLD) measurements were used to investigate dose uniformity across field junctions as a function of field separation and, in particular, to review the CCSG recommendation for the treatment of medulloblastoma with separate head and spine fields.

  12. Evaluation of prompt nucleation of bubbles in annular fuel elements during the initial depressurization transient of a DEGB LOCA

    SciTech Connect

    Smith, A.C.

    1997-06-01

    In the first moments following the pipe break, of a DEGB LOCA, the depressurization wave is postulated to propagate rapidly through the system, in the manner of an acoustic or water hammer wave. this is immediately followed by a (reflected) repressurization wave, as the flow of coolant through the break is established. The pressure history is then dictated by the flow from the break and the ability of the pressurizer, pumps and accumulators to supply coolant. The initial sudden drop in pressure may result in the system pressure falling below the saturation pressure of the coolant. This could, in turn, result in bubble formation. Such immediate vapor formation (prompt nucleation of bubbles), in the period before the repressurization wave restores the system pressure to a level above the saturation pressure might initiate flow instability. Such an interruption in flow would allow the fuel tube clad temperature to increase rapidly. Depending on the duration of the flow interruption, the reactor might not be able to survive the initial moments of DEGB LOCA. It has generally been that this phenomenon would not actually occur in an operating reactor. The purpose of this investigation is to evaluate the possibility of occurrence of bubble formation as a result of initial depressurization. 7 refs., 6 figs.

  13. Electrolytic reduction of a simulated oxide spent fuel and the fates of representative elements in a Li2O-LiCl molten salt

    NASA Astrophysics Data System (ADS)

    Park, Wooshin; Choi, Eun-Young; Kim, Sung-Wook; Jeon, Sang-Chae; Cho, Young-Hwan; Hur, Jin-Mok

    2016-08-01

    A series of electrolytic reduction experiments were carried out using a simulated oxide spent fuel to investigate the reduction behavior of elements in a mixed oxide condition and the fates of elements in the reduction process with 1.0 wt% Li2O-LiCl. It was found out that 155% of the theoretical charge was enough to reduce the simulated. Te and Eu were expected to possibly exist in the precipitate and on the anode surface, whereas Ba and Sr showed apparent dissolution behaviors. Rare earths showed relatively low metal fractions from 28.2 to 34.0% except for Y. And the solubility of rare earths was observed to be low due to the low concentration of Li2O. The reduction of U was successful as expected showing 99.8% of a metal fraction. Also it was shown that the reduction of ZrO2 would be effective when a relatively small amount was included in a metal oxide mixture.

  14. Turbine combustor with fuel nozzles having inner and outer fuel circuits

    DOEpatents

    Uhm, Jong Ho; Johnson, Thomas Edward; Kim, Kwanwoo

    2013-12-24

    A combustor cap assembly for a turbine engine includes a combustor cap and a plurality of fuel nozzles mounted on the combustor cap. One or more of the fuel nozzles would include two separate fuel circuits which are individually controllable. The combustor cap assembly would be controlled so that individual fuel circuits of the fuel nozzles are operated or deliberately shut off to provide for physical separation between the flow of fuel delivered by adjacent fuel nozzles and/or so that adjacent fuel nozzles operate at different pressure differentials. Operating a combustor cap assembly in this fashion helps to reduce or eliminate the generation of undesirable and potentially harmful noise.

  15. Impact of Oxy-Fuel Conditions on Elemental Mercury Re-Emission in Wet Flue Gas Desulfurization Systems.

    PubMed

    Fernández-Miranda, Nuria; Lopez-Anton, M Antonia; Torre-Santos, Teresa; Díaz-Somoano, Mercedes; Martínez-Tarazona, M Rosa

    2016-07-01

    This study evaluates some of the variables that may influence mercury retention in wet flue gas desulfurization (WFGD) plants, focusing on oxy-coal combustion processes and differences when compared with atmospheres enriched in N2. The main drawback of using WFGD for mercury capture is the possibility of unwanted reduction of dissolved Hg(2+), leading to the re-emission of insoluble elemental mercury (Hg(0)), which decreases efficiency. To acquire a better understanding of the mercury re-emission reactions in WFGD systems, this work analyses different variables that influence the behavior of mercury in slurries obtained from two limestones, under an oxy-combustion atmosphere. The O2 supplied to the reactor, the influence of the pH, the concentration of mercury in the gas phase, and the enhancement of mercury in the slurry were the variables considered. The study was performed at laboratory scale, where possible reactions between the components in the scrubber can be individually evaluated. It was found that in an oxy-combustion atmosphere (mostly CO2), the re-emission of Hg(0) is lower than under a N2-enriched atmosphere, and the mercury is mainly retained as Hg(2+) in the liquid phase. PMID:27329988

  16. Adjacent Segment Pathology after Lumbar Spinal Fusion.

    PubMed

    Lee, Jae Chul; Choi, Sung-Woo

    2015-10-01

    One of the major clinical issues encountered after lumbar spinal fusion is the development of adjacent segment pathology (ASP) caused by increased mechanical stress at adjacent segments, and resulting in various radiographic changes and clinical symptoms. This condition may require surgical intervention. The incidence of ASP varies with both the definition and methodology adopted in individual studies; various risk factors for this condition have been identified, although a significant controversy still exists regarding their significance. Motion-preserving devices have been developed, and some studies have shown their efficacy of preventing ASP. Surgeons should be aware of the risk factors of ASP when planning a surgery, and accordingly counsel their patients preoperatively. PMID:26435804

  17. Adjacent Segment Pathology after Anterior Cervical Fusion

    PubMed Central

    Chung, Jae Yoon; Park, Jong-Beom; Seo, Hyoung-Yeon

    2016-01-01

    Anterior cervical fusion has become a standard of care for numerous pathologic conditions of the cervical spine. However, subsequent development of clinically significant disc disease at levels adjacent to fused discs is a serious long-term complication of this procedure. As more patients live longer after surgery, it is foreseeable that adjacent segment pathology (ASP) will develop in increasing numbers of patients. Also, ASP has been studied more intensively with the recent popularity of motion preservation technologies like total disc arthroplasty. The true nature and scope of ASP remains poorly understood. The etiology of ASP is most likely multifactorial. Various factors including altered biomechanical stresses, surgical disruption of soft tissue and the natural history of cervical disc disease contribute to the development of ASP. General factors associated with disc degeneration including gender, age, smoking and sports may play a role in the development of ASP. Postoperative sagittal alignment and type of surgery are also considered potential causes of ASP. Therefore, a spine surgeon must be particularly careful to avoid unnecessary disruption of the musculoligamentous structures, reduced risk of direct injury to the disc during dissection and maintain a safe margin between the plate edge and adjacent vertebrae during anterior cervical fusion. PMID:27340541

  18. Adjacent Segment Pathology after Anterior Cervical Fusion.

    PubMed

    Chung, Jae Yoon; Park, Jong-Beom; Seo, Hyoung-Yeon; Kim, Sung Kyu

    2016-06-01

    Anterior cervical fusion has become a standard of care for numerous pathologic conditions of the cervical spine. However, subsequent development of clinically significant disc disease at levels adjacent to fused discs is a serious long-term complication of this procedure. As more patients live longer after surgery, it is foreseeable that adjacent segment pathology (ASP) will develop in increasing numbers of patients. Also, ASP has been studied more intensively with the recent popularity of motion preservation technologies like total disc arthroplasty. The true nature and scope of ASP remains poorly understood. The etiology of ASP is most likely multifactorial. Various factors including altered biomechanical stresses, surgical disruption of soft tissue and the natural history of cervical disc disease contribute to the development of ASP. General factors associated with disc degeneration including gender, age, smoking and sports may play a role in the development of ASP. Postoperative sagittal alignment and type of surgery are also considered potential causes of ASP. Therefore, a spine surgeon must be particularly careful to avoid unnecessary disruption of the musculoligamentous structures, reduced risk of direct injury to the disc during dissection and maintain a safe margin between the plate edge and adjacent vertebrae during anterior cervical fusion. PMID:27340541

  19. Impact of adjacent land use on coastal wetland sediments.

    PubMed

    Karstens, Svenja; Buczko, Uwe; Jurasinski, Gerald; Peticzka, Robert; Glatzel, Stephan

    2016-04-15

    Coastal wetlands link terrestrial with marine ecosystems and are influenced from both land and sea. Therefore, they are ecotones with strong biogeochemical gradients. We analyzed sediment characteristics including macronutrients (C, N, P, K, Mg, Ca, S) and heavy metals (Mn, Fe, Cu, Zn, Al, Co, Cr, Ni) of two coastal wetlands dominated by Phragmites australis at the Darss-Zingst Bodden Chain, a lagoon system at the Southern Baltic Sea, to identify the impact of adjacent land use and to distinguish between influences from land or sea. In the wetland directly adjacent to cropland (study site Dabitz) heavy metal concentrations were significantly elevated. Fertilizer application led to heavy metal accumulation in the sediments of the adjacent wetland zones. In contrast, at the other study site (Michaelsdorf), where the hinterland has been used as pasture, heavy metal concentrations were low. While the amount of macronutrients was also influenced by vegetation characteristics (e.g. carbon) or water chemistry (e.g. sulfate), the accumulation of heavy metals is regarded as purely anthropogenic influence. A principal component analysis (PCA) based on the sediment data showed that the wetland fringes of the two study sites are not distinguishable, neither in their macronutrient status nor in their concentrations of heavy metals, whereas the interior zones exhibit large differences in terms of heavy metal concentrations. This suggests that seaside influences are minor compared to influences from land. Altogether, heavy metal concentrations were still below national precautionary and action values. However, if we regard the macronutrient and heavy metal concentrations in the wetland fringes as the natural background values, an accumulation of trace elements from agricultural production in the hinterland is apparent. Thus, coastal wetlands bordering croplands may function as effective pollutant buffers today, but the future development has to be monitored closely to avoid

  20. Fuel cell water transport

    DOEpatents

    Vanderborgh, Nicholas E.; Hedstrom, James C.

    1990-01-01

    The moisture content and temperature of hydrogen and oxygen gases is regulated throughout traverse of the gases in a fuel cell incorporating a solid polymer membrane. At least one of the gases traverses a first flow field adjacent the solid polymer membrane, where chemical reactions occur to generate an electrical current. A second flow field is located sequential with the first flow field and incorporates a membrane for effective water transport. A control fluid is then circulated adjacent the second membrane on the face opposite the fuel cell gas wherein moisture is either transported from the control fluid to humidify a fuel gas, e.g., hydrogen, or to the control fluid to prevent excess water buildup in the oxidizer gas, e.g., oxygen. Evaporation of water into the control gas and the control gas temperature act to control the fuel cell gas temperatures throughout the traverse of the fuel cell by the gases.

  1. Fuel cell system with interconnect

    SciTech Connect

    Liu, Zhien; Goettler, Richard

    2015-09-29

    The present invention includes a fuel cell system having a plurality of adjacent electrochemical cells formed of an anode layer, a cathode layer spaced apart from the anode layer, and an electrolyte layer disposed between the anode layer and the cathode layer. The fuel cell system also includes at least one interconnect, the interconnect being structured to conduct free electrons between adjacent electrochemical cells. Each interconnect includes a primary conductor embedded within the electrolyte layer and structured to conduct the free electrons.

  2. Fuel cell system with interconnect

    SciTech Connect

    Goettler, Richard; Liu, Zhien

    2015-08-11

    The present invention includes a fuel cell system having a plurality of adjacent electrochemical cells formed of an anode layer, a cathode layer spaced apart from the anode layer, and an electrolyte layer disposed between the anode layer and the cathode layer. The fuel cell system also includes at least one interconnect, the interconnect being structured to conduct free electrons between adjacent electrochemical cells. Each interconnect includes a primary conductor embedded within the electrolyte layer and structured to conduct the free electrons.

  3. Fuel cell system with interconnect

    SciTech Connect

    Goettler, Richard; Liu, Zhien

    2015-03-10

    The present invention includes a fuel cell system having a plurality of adjacent electrochemical cells formed of an anode layer, a cathode layer spaced apart from the anode layer, and an electrolyte layer disposed between the anode layer and the cathode layer. The fuel cell system also includes at least one interconnect, the interconnect being structured to conduct free electrons between adjacent electrochemical cells. Each interconnect includes a primary conductor embedded within the electrolyte layer and structured to conduct the free electrons.

  4. Features of temperature control of fuel element cladding for pressurized water nuclear reactor “WWER-1000” while simulating reactor accidents

    SciTech Connect

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M.

    2013-09-11

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones “cladding-junction” and “junction-coolant”. The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ∼5% of maximum value by the final moment of the stage of linear increase in the temperature.

  5. Reconstructing genome mixtures from partial adjacencies.

    PubMed

    Mahmoody, Ahmad; Kahn, Crystal L; Raphael, Benjamin J

    2012-01-01

    Many cancer genome sequencing efforts are underway with the goal of identifying the somatic mutations that drive cancer progression. A major difficulty in these studies is that tumors are typically heterogeneous, with individual cells in a tumor having different complements of somatic mutations. However, nearly all DNA sequencing technologies sequence DNA from multiple cells, thus resulting in measurement of mutations from a mixture of genomes. Genome rearrangements are a major class of somatic mutations in many tumors, and the novel adjacencies (i.e. breakpoints) resulting from these rearrangements are readily detected from DNA sequencing reads. However, the assignment of each rearrangement, or adjacency, to an individual cancer genome in the mixture is not known. Moreover, the quantity of DNA sequence reads may be insufficient to measure all rearrangements in all genomes in the tumor. Motivated by this application, we formulate the k-minimum completion problem (k-MCP). In this problem, we aim to reconstruct k genomes derived from a single reference genome, given partial information about the adjacencies present in the mixture of these genomes. We show that the 1-MCP is solvable in linear time in the cases where: (i) the measured, incomplete genome has a single circular or linear chromosome; (ii) there are no restrictions on the chromosomal content of the measured, incomplete genome. We also show that the k-MCP problem, for k ≥ 3 in general, and the 2-MCP problem with the double-cut-and-join (DCJ) distance are NP-complete, when there are no restriction on the chromosomal structure of the measured, incomplete genome. These results lay the foundation for future algorithmic studies of the k-MCP and the application of these algorithms to real cancer sequencing data. PMID:23282028

  6. Positive adjacency effects mediated by seed disperser birds in pine plantations.

    PubMed

    Zamora, Regino; Hódar, José Antonio; Matías, Luís; Mendoza, Irene

    2010-06-01

    This study examines the consequences of adjacent elements for a given patch, through their effects on zoochorous dispersion by frugivorous birds. The case study consists of pine plantations (the focal patch) adjacent to other patches of native vegetation (mixed patches of native forest and shrublands), and/or pine plantations. Our hypothesis is that input of native woody species propagules generated by frugivorous birds within plantations strongly depends on the nature of the surrounding vegetation. To test this hypothesis, we studied frugivorous-bird abundance, seed dispersion, and seedling establishment in nine pine plantation plots in contact with patches of native vegetation. To quantify adjacency arrangement effects, we used the percentage of common border between a patch and each of its adjacent elements. Frugivorous bird occurrence in pine plantations is influenced by the adjacent vegetation: the greater the contact with native vegetation patches, the more abundant were the frugivorous birds within pine plantations. Furthermore, frugivorous birds introduce into plantations the seeds of a large sample of native fleshy-fruited species. The results confirm the hypothesis that zoochorous seed rain is strongly determined by the kind of vegetation surrounding a given plantation. This finding underlines the importance of the composition of the mosaic surrounding plantations and the availability of mobile link species as key landscape features conditioning passive restoration processes. PMID:20597289

  7. Fuel cell generator energy dissipator

    DOEpatents

    Veyo, Stephen Emery; Dederer, Jeffrey Todd; Gordon, John Thomas; Shockling, Larry Anthony

    2000-01-01

    An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a fuel cell generator when the electrical power output of the fuel cell generator is terminated. During a generator shut down condition, electrically resistive elements are automatically connected across the fuel cell generator terminals in order to draw current, thereby depleting the fuel

  8. NEUTRONIC REACTOR CONTROL ELEMENT

    DOEpatents

    Newson, H.W.

    1960-09-13

    A novel composite neutronic reactor control element is offered. The element comprises a multiplicity of sections arranged in end-to-end relationship, each of the sections having a markedly different neutron-reactive characteristic. For example, a three-section control element could contain absorber, moderator, and fuel sections. By moving such an element longitudinally through a reactor core, reactivity is decreased by the absorber, increased slightly by the moderator, or increased substantially by the fuel. Thus, control over a wide reactivity range is provided.

  9. Fuel cell arrangement

    DOEpatents

    Isenberg, Arnold O.

    1987-05-12

    A fuel cell arrangement is provided wherein cylindrical cells of the solid oxide electrolyte type are arranged in planar arrays where the cells within a plane are parallel. Planes of cells are stacked with cells of adjacent planes perpendicular to one another. Air is provided to the interior of the cells through feed tubes which pass through a preheat chamber. Fuel is provided to the fuel cells through a channel in the center of the cell stack; the fuel then passes the exterior of the cells and combines with the oxygen-depleted air in the preheat chamber.

  10. Fuel cell arrangement

    DOEpatents

    Isenberg, A.O.

    1987-05-12

    A fuel cell arrangement is provided wherein cylindrical cells of the solid oxide electrolyte type are arranged in planar arrays where the cells within a plane are parallel. Planes of cells are stacked with cells of adjacent planes perpendicular to one another. Air is provided to the interior of the cells through feed tubes which pass through a preheat chamber. Fuel is provided to the fuel cells through a channel in the center of the cell stack; the fuel then passes the exterior of the cells and combines with the oxygen-depleted air in the preheat chamber. 3 figs.

  11. METHOD FOR MAKING FUEL ELEMENTS

    DOEpatents

    Kates, L.W.; Campbell, R.W.; Heartel, R.H.W.

    1960-08-01

    A method is given for making zirconium-clad uranium wire. A tube of zirconium is closed with a zirconium plug, after which a chilled uranium core is inserted in the tube to rest against the plug. Additional plugs and cores are inserted alternately as desired. The assembly is then sheathed with iron, hot worked to the desired size, and the iron sheath removed.

  12. REGENERATION OF REACTOR FUEL ELEMENTS

    DOEpatents

    Lyon, W.L.

    1960-04-01

    A process is described for concentrating uranium and/or plutonium metal in aluminum alloys in which the actinide content was partially consumed by neutron bombardinent. Two embodiments are claimed: Either the alloy is heated, together with zinc chloride to at least 1000 deg C whereby some aluminum, in the form of aluminum chloride, and any zinc formed volatilize; or else aluminum fluoride is added and reacted at 800 to 1000 deg O and substmospheric pressure whereby pant of the aluminum volatilizes and aluminum subfluoride.

  13. REGENERATION OF REACTOR FUEL ELEMENTS

    DOEpatents

    Roake, W.E.; Lyon, W.L.

    1960-03-29

    A process of concentrating by electrolysis the uraatum and/or plutonium of an aluminum alloy containing these actinides after the actinide has been partially consumed by neutron bombardment in a reactor is given. The alloy is made the anode in a system having an aluminum cathode and a cryolite electrolyte. Electrolysis from 22 to 28 ampere-hours removes a sufficient quantity of aluminum from the alloy to make it suitable for reuse.

  14. 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.

  15. Sound insulation property of membrane-type acoustic metamaterials carrying different masses at adjacent cells

    NASA Astrophysics Data System (ADS)

    Zhang, Yuguang; Wen, Jihong; Zhao, Honggang; Yu, Dianlong; Cai, Li; Wen, Xisen

    2013-08-01

    We present the experimental realization and theoretical understanding of membrane-type acoustic metamaterials embedded with different masses at adjacent cells, capable of increasing the transmission loss at low frequency. Owing to the reverse vibration of adjacent cells, Transmission loss (TL) peaks appear, and the magnitudes of the TL peaks exceed the predicted results of the composite wall. Compared with commonly used configuration, i.e., all cells carrying with identical mass, the nonuniformity of attaching masses causes another much low TL peak. Finite element analysis was employed to validate and provide insights into the TL behavior of the structure.

  16. Exchange coupling between laterally adjacent nanomagnets.

    PubMed

    Dey, H; Csaba, G; Bernstein, G H; Porod, W

    2016-09-30

    We experimentally demonstrate exchange-coupling between laterally adjacent nanomagnets. Our results show that two neighboring nanomagnets that are each antiferromagnetically exchange-coupled to a common ferromagnetic bottom layer can be brought into strong ferromagnetic interaction. Simulations show that interlayer exchange coupling effectively promotes ferromagnetic alignment between the two nanomagnets, as opposed to antiferromagnetic alignment due to dipole-coupling. In order to experimentally demonstrate the proposed scheme, we fabricated arrays of pairs of elongated, single-domain nanomagnets. Magnetic force microscopy measurements show that most of the pairs are ferromagnetically ordered. The results are in agreement with micromagnetic simulations. The presented scheme can achieve coupling strengths that are significantly stronger than dipole coupling, potentially enabling far-reaching applications in Nanomagnet Logic, spin-wave devices and three-dimensional storage and computing. PMID:27535227

  17. Seismicity in Azerbaijan and Adjacent Caspian Sea

    SciTech Connect

    Panahi, Behrouz M.

    2006-03-23

    So far no general view on the geodynamic evolution of the Black Sea to the Caspian Sea region is elaborated. This is associated with the geological and structural complexities of the region revealed by geophysical, geochemical, petrologic, structural, and other studies. A clash of opinions on geodynamic conditions of the Caucasus region, sometimes mutually exclusive, can be explained by a simplified interpretation of the seismic data. In this paper I analyze available data on earthquake occurrences in Azerbaijan and the adjacent Caspian Sea region. The results of the analysis of macroseismic and instrumental data, seismic regime, and earthquake reoccurrence indicate that a level of seismicity in the region is moderate, and seismic event are concentrated in the shallow part of the lithosphere. Seismicity is mostly intra-plate, and spatial distribution of earthquake epicenters does not correlate with the plate boundaries.

  18. Boundary Layers of Air Adjacent to Cylinders

    PubMed Central

    Nobel, Park S.

    1974-01-01

    Using existing heat transfer data, a relatively simple expression was developed for estimating the effective thickness of the boundary layer of air surrounding cylinders. For wind velocities from 10 to 1000 cm/second, the calculated boundary-layer thickness agreed with that determined for water vapor diffusion from a moistened cylindrical surface 2 cm in diameter. It correctly predicted the resistance for water vapor movement across the boundary layers adjacent to the (cylindrical) inflorescence stems of Xanthorrhoea australis R. Br. and Scirpus validus Vahl and the leaves of Allium cepa L. The boundary-layer thickness decreased as the turbulence intensity increased. For a turbulence intensity representative of field conditions (0.5) and for νwindd between 200 and 30,000 cm2/second (where νwind is the mean wind velocity and d is the cylinder diameter), the effective boundary-layer thickness in centimeters was equal to [Formula: see text]. PMID:16658855

  19. Differences between the oxidation behaviour of A3 fuel element matrix graphites in air and in steam and its relevance on accident progress in HTRs

    SciTech Connect

    Kuehn, K.; Hinssen, H-K.; Moormann, R.

    2004-07-01

    The fuel element matrix graphites A3-3 and A3-27 were used in High Temperature Reactor fuel pebbles for many years. However, these materials show as other graphites a limited oxidation resistance in contact to oxidising gases (air and steam), which even decreases with increasing temperatures: In HTRs, having in normal operation a non-oxidising environment, an ingress of air or steam leads to corrosion of graphite with the potential of enhanced fission product release Matrix graphites differ by its coked binder content from standard nuclear graphites (e.g. V483T, ASR-1RG, IG110, H-451), where both filler and binder are completely graphitized. The influence of this coked binder content on the oxidation behaviour will be discussed in this paper. Experimental investigations with A3-3 and A3-27 were done in oxygen (air) at temperatures between 673 - 1023 K and in steam between 1173 - 1253 K. These experiments took place under isothermal conditions in the chemical regime, where the chemical reaction itself is the rate limiting step and a homogeneous oxidation inside of the sample occurs. The experiments reveal different oxidation behaviour as well between binder and filler component as between oxidation behaviour in oxygen (air) and steam. In air at low temperatures two rate maxima are observed: The first maximum attend in all experiments at {approx} 5 % burn off, a second one at higher burn off values (35 - 45 % burn off). These rate maxima can be explained by a selective binder-filler oxidation: The first peak at 5 % burn off is due to the oxidation of the binder, the second peak at higher burn off values is caused by the oxidation of the remaining filler. At higher temperatures in air the filler peak becomes more pronounced and the binder peak vanishes, which is due to the lower activation energy of binder oxidation compared to the filler. In steam this behaviour appears contrary: A maximum at 5 % burn off, which is probably also connected to the binder, is observed

  20. Sensor system for fuel transport vehicle

    DOEpatents

    Earl, Dennis Duncan; McIntyre, Timothy J.; West, David L.

    2016-03-22

    An exemplary sensor system for a fuel transport vehicle can comprise a fuel marker sensor positioned between a fuel storage chamber of the vehicle and an access valve for the fuel storage chamber of the vehicle. The fuel marker sensor can be configured to measure one or more characteristics of one or more fuel markers present in the fuel adjacent the sensor, such as when the marked fuel is unloaded at a retail station. The one or more characteristics can comprise concentration and/or identity of the one or more fuel markers in the fuel. Based on the measured characteristics of the one or more fuel markers, the sensor system can identify the fuel and/or can determine whether the fuel has been adulterated after the marked fuel was last measured, such as when the marked fuel was loaded into the vehicle.

  1. 30 CFR 56.9103 - Clearance on adjacent tracks.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Clearance on adjacent tracks. 56.9103 Section..., Hauling, and Dumping Traffic Safety § 56.9103 Clearance on adjacent tracks. Railcars shall not be left on side tracks unless clearance is provided for traffic on adjacent tracks....

  2. 30 CFR 57.9103 - Clearance on adjacent tracks.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 30 Mineral Resources 1 2011-07-01 2011-07-01 false Clearance on adjacent tracks. 57.9103 Section..., Hauling, and Dumping Traffic Safety § 57.9103 Clearance on adjacent tracks. Railcars shall not be left on side tracks unless clearance is provided for traffic on adjacent tracks....

  3. 30 CFR 56.9103 - Clearance on adjacent tracks.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 30 Mineral Resources 1 2011-07-01 2011-07-01 false Clearance on adjacent tracks. 56.9103 Section..., Hauling, and Dumping Traffic Safety § 56.9103 Clearance on adjacent tracks. Railcars shall not be left on side tracks unless clearance is provided for traffic on adjacent tracks....

  4. 30 CFR 57.9103 - Clearance on adjacent tracks.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Clearance on adjacent tracks. 57.9103 Section..., Hauling, and Dumping Traffic Safety § 57.9103 Clearance on adjacent tracks. Railcars shall not be left on side tracks unless clearance is provided for traffic on adjacent tracks....

  5. 33 CFR 80.1395 - Puget Sound and adjacent waters.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Puget Sound and adjacent waters... INTERNATIONAL NAVIGATION RULES COLREGS DEMARCATION LINES Thirteenth District § 80.1395 Puget Sound and adjacent waters. The 72 COLREGS shall apply on all waters of Puget Sound and adjacent waters, including Lake...

  6. 33 CFR 80.1395 - Puget Sound and adjacent waters.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 33 Navigation and Navigable Waters 1 2013-07-01 2013-07-01 false Puget Sound and adjacent waters... INTERNATIONAL NAVIGATION RULES COLREGS DEMARCATION LINES Thirteenth District § 80.1395 Puget Sound and adjacent waters. The 72 COLREGS shall apply on all waters of Puget Sound and adjacent waters, including Lake...

  7. 33 CFR 80.1395 - Puget Sound and adjacent waters.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 33 Navigation and Navigable Waters 1 2011-07-01 2011-07-01 false Puget Sound and adjacent waters... INTERNATIONAL NAVIGATION RULES COLREGS DEMARCATION LINES Thirteenth District § 80.1395 Puget Sound and adjacent waters. The 72 COLREGS shall apply on all waters of Puget Sound and adjacent waters, including Lake...

  8. 33 CFR 80.1395 - Puget Sound and adjacent waters.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 33 Navigation and Navigable Waters 1 2014-07-01 2014-07-01 false Puget Sound and adjacent waters... INTERNATIONAL NAVIGATION RULES COLREGS DEMARCATION LINES Thirteenth District § 80.1395 Puget Sound and adjacent waters. The 72 COLREGS shall apply on all waters of Puget Sound and adjacent waters, including Lake...

  9. 33 CFR 80.1395 - Puget Sound and adjacent waters.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 33 Navigation and Navigable Waters 1 2012-07-01 2012-07-01 false Puget Sound and adjacent waters... INTERNATIONAL NAVIGATION RULES COLREGS DEMARCATION LINES Thirteenth District § 80.1395 Puget Sound and adjacent waters. The 72 COLREGS shall apply on all waters of Puget Sound and adjacent waters, including Lake...

  10. Fuel-cycle cost comparisons with oxide and silicide fuels

    SciTech Connect

    Matos, J.E.; Freese, K.E.

    1982-01-01

    This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed.

  11. Axially staggered seed-blanket reactor fuel module construction

    DOEpatents

    Cowell, Gary K.; DiGuiseppe, Carl P.

    1985-01-01

    A heterogeneous nuclear reactor of the seed-blanket type is provided wher the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements. The arrangements of the fissile and fertile regions in an alternating axial manner minimizes the radial power peaking factors and provides a more optional thermal-hydraulic design than is afforded by radial arrangements.

  12. Interaction between adjacent lightning discharges in clouds

    NASA Astrophysics Data System (ADS)

    Wang, Yanhui; Zhang, Guangshu; Zhang, Tong; Li, Yajun; Wu, Bin; Zhang, Tinglong

    2013-07-01

    Using a 3D lightning radiation source locating system (LLS), three pairs of associated lightning discharges (two or more adjacent lightning discharges following an arbitrary rule that their space-gap was less than 10 km and their time-gap was less than 800 ms) were observed, and the interaction between associated lightning discharges was analyzed. All these three pairs of associated lightning discharges were found to involve three or more charge regions (the ground was considered as a special charge region). Moreover, at least one charge region involved two lightning discharges per pair of associated lightning discharges. Identified from electric field changes, the subsequent lightning discharges were suppressed by the prior lightning discharges. However, it is possible that the prior lightning discharge provided a remaining discharge channel to facilitate the subsequent lightning discharge. The third case provided evidence of this possibility. Together, the results suggested that, if the charges in the main negative charge region can be consumed using artificial lightning above the main negative charge regions, lightning accidents on the ground could be greatly reduced, on the condition that the height of the main negative charge region and the charge intensity of the lower positive charge region are suitable.

  13. The relative influence of road characteristics and habitat on adjacent lizard populations in arid shrublands

    USGS Publications Warehouse

    Hubbard, Kaylan A.; Chalfoun, Anna L.; Gerow, Kenneth G.

    2016-01-01

    As road networks continue to expand globally, indirect impacts to adjacent wildlife populations remain largely unknown. Simultaneously, reptile populations are declining worldwide and anthropogenic habitat loss and fragmentation are frequently cited causes. We evaluated the relative influence of three different road characteristics (surface treatment, width, and traffic volume) and habitat features on adjacent populations of Northern Sagebrush Lizards (Sceloporus graciosus graciosus), Plateau Fence Lizards (S. tristichus), and Greater Short-Horned Lizards (Phrynosoma hernandesi) in mixed arid shrubland habitats in southwest Wyoming. Neither odds of lizard presence nor relative abundance was significantly related to any of the assessed road characteristics, although there was a trend for higher Sceloporus spp. abundance adjacent to paved roads. Sceloporus spp. relative abundance did not vary systematically with distance to the nearest road. Rather, both Sceloporus spp. and Greater Short-Horned Lizards were associated strongly with particular habitat characteristics adjacent to roads. Sceloporus spp. presence and relative abundance increased with rock cover, relative abundance was associated positively with shrub cover, and presence was associated negatively with grass cover. Greater Short-Horned Lizard presence increased with bare ground and decreased marginally with shrub cover. Our results suggest that habitat attributes are stronger correlates of lizard presence and relative abundance than individual characteristics of adjacent roads, at least in our system. Therefore, an effective conservation approach for these species may be to consider the landscape through which new roads and their associated development would occur, and the impact that placement could have on fragment size and key habitat elements.

  14. Fuel flexible fuel injector

    SciTech Connect

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  15. Coordinated tissue-specific regulation of adjacent alternative 3′ splice sites in C. elegans

    PubMed Central

    Ragle, James Matthew; Katzman, Sol; Akers, Taylor F.; Barberan-Soler, Sergio; Zahler, Alan M.

    2015-01-01

    Adjacent alternative 3′ splice sites, those separated by ≤18 nucleotides, provide a unique problem in the study of alternative splicing regulation; there is overlap of the cis-elements that define the adjacent sites. Identification of the intron's 3′ end depends upon sequence elements that define the branchpoint, polypyrimidine tract, and terminal AG dinucleotide. Starting with RNA-seq data from germline-enriched and somatic cell-enriched Caenorhabditis elegans samples, we identify hundreds of introns with adjacent alternative 3′ splice sites. We identify 203 events that undergo tissue-specific alternative splicing. For these, the regulation is monodirectional, with somatic cells preferring to splice at the distal 3′ splice site (furthest from the 5′ end of the intron) and germline cells showing a distinct shift toward usage of the adjacent proximal 3′ splice site (closer to the 5′ end of the intron). Splicing patterns in somatic cells follow C. elegans consensus rules of 3′ splice site definition; a short stretch of pyrimidines preceding an AG dinucleotide. Splicing in germline cells occurs at proximal 3′ splice sites that lack a preceding polypyrimidine tract, and in three instances the germline-specific site lacks the AG dinucleotide. We provide evidence that use of germline-specific proximal 3′ splice sites is conserved across Caenorhabditis species. We propose that there are differences between germline and somatic cells in the way that the basal splicing machinery functions to determine the intron terminus. PMID:25922281

  16. Fuel pumping system and method

    DOEpatents

    Shafer, Scott F.; Wang, Lifeng ,

    2006-12-19

    A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

  17. Fuel Pumping System And Method

    DOEpatents

    Shafer, Scott F.; Wang, Lifeng

    2005-12-13

    A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

  18. Petroleum basins of Sakhalin and adjacent shelf

    SciTech Connect

    Mavrinski, Y.; Koblov, E. )

    1993-09-01

    Sixty-seven oil and gas fields have been discovered on Sakhalin and the adjacent shelf but the distribution of fields is uneven in north Sakhalin, south Sakhalin, and the Tatar basins. The sedimentary cover is composed of sandy, clayey, and siliceous rocks, with volcanogenic and coal-bearing deposits of Upper Cretaceous, Paleogene, and Neogene 8-12 km thick. Marine clayey and siliceous oil source rocks are regionally developed in the section at different stratigraphic levels; the organic matter is of mixed type and the content varies from 0.5 to 1.5%. The upper Oligocene and middle-upper Miocene source rocks in the north Sakhalin basin are typical, and the organic carbon content ranges from 1 to 5%. The level of organic matter catagenesis and conversion into hydrocarbons is high because of the high differential geothermal gradient in the basins, 30-50[degrees]C per km. Porous sandstones in the Miocene form the reservoirs in all fields with the exception of Okruzhnoye, where the pay zone is a siliceous claystone. Growth-fault rollovers and anticlines form the main traps ranging in area from 5 to 300 km[sup 2], with amplitudes between 100 and 600 m. both stratigraphic and structural traps have been identified. Considerable volumes of reserves are associated with the Miocene deposits of north Sakhalin, which are characterized by an optimum combination of oil source rocks, focused migration paths, and thick sequences of reservoirs and cap rocks. Six large fields have been discovered in the past 15 yr. Oil and condensate reserves stand at over 300 million MT, and gas reserves are about 900 billion m[sup 3].

  19. Apparatus and method for grounding compressed fuel fueling operator

    DOEpatents

    Cohen, Joseph Perry; Farese, David John; Xu, Jianguo

    2002-06-11

    A safety system for grounding an operator at a fueling station prior to removing a fuel fill nozzle from a fuel tank upon completion of a fuel filling operation is provided which includes a fuel tank port in communication with the fuel tank for receiving and retaining the nozzle during the fuel filling operation and a grounding device adjacent to the fuel tank port which includes a grounding switch having a contact member that receives physical contact by the operator and where physical contact of the contact member activates the grounding switch. A releasable interlock is included that provides a lock position wherein the nozzle is locked into the port upon insertion of the nozzle into the port and a release position wherein the nozzle is releasable from the port upon completion of the fuel filling operation and after physical contact of the contact member is accomplished.

  20. Fuel dissipater for pressurized fuel cell generators

    DOEpatents

    Basel, Richard A.; King, John E.

    2003-11-04

    An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a pressurized fuel cell generator (10) when the electrical power output of the fuel cell generator is terminated during transient operation, such as a shutdown; where, two electrically resistive elements (two of 28, 53, 54, 55) at least one of which is connected in parallel, in association with contactors (26, 57, 58, 59), a multi-point settable sensor relay (23) and a circuit breaker (24), are automatically connected across the fuel cell generator terminals (21, 22) at two or more contact points, in order to draw current, thereby depleting the fuel inventory in the generator.

  1. Bipolar fuel cell

    DOEpatents

    McElroy, James F.

    1989-01-01

    The present invention discloses an improved fuel cell utilizing an ion transporting membrane having a catalytic anode and a catalytic cathode bonded to opposite sides of the membrane, a wet-proofed carbon sheet in contact with the cathode surface opposite that bonded to the membrane and a bipolar separator positioned in electrical contact with the carbon sheet and the anode of the adjacent fuel cell. Said bipolar separator and carbon sheet forming an oxidant flowpath, wherein the improvement comprises an electrically conductive screen between and in contact with the wet-proofed carbon sheet and the bipolar separator improving the product water removal system of the fuel cell.

  2. Design package for fuel retrieval system fuel handling tool modification

    SciTech Connect

    TEDESCHI, D.J.

    1999-03-17

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports.

  3. Design Package for Fuel Retrieval System Fuel Handling Tool Modification

    SciTech Connect

    TEDESCHI, D.J.

    2000-03-27

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports.

  4. The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors: A preliminary assessment of experiments HRB-17, HFR-B1, HFR-K6 and KORA

    SciTech Connect

    Myers, B.F.

    1995-09-01

    The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors has been measured in different laboratories under both irradiation and post irradiation conditions. The data from experiments HRB-17, HFR-B1, HFR-K6, and in the KORA facility are compared to assess their consistency and complimentarily. The experiments are consistent under comparable experimental conditions and reveal two general mechanisms involving exposed fuel kernels embedded in carbonaceous materials. One is manifest as a strong dependence of fission gas release on the partial pressure of water vapor below 1 kPa and the other, as a weak dependence above 1 kPa.

  5. Contents, seasonal variations, and forms of migration of major and minor elements in surface waters in the area of the Tyrnyauz Tungsten-Molybdenum Combine (TTMC) and adjacent areas (Kabardino-Balkarian Republic, Russian Federation) and actions for recovery of the ecological environment

    NASA Astrophysics Data System (ADS)

    Vinokurov, S. F.; Gurbanov, A. G.; Bogatikov, O. A.; Karamurzov, B. S.; Gazeev, V. M.; Lexin, A. B.; Shevchenko, A. V.; Dolov, S. M.; Dudarov, Z. I.

    2016-04-01

    Anomalous concentrations of numerous major and minor elements significantly exceeding the threshold limit values (TLV) for drinking water were registered in the area of the Tyrnyauz Tungsten-Molybdenum Combine (TTMC). The maximal excess of the TLV (by one or two orders of magnitude) were obtained for Mo (up to 11 mg/L), W (4.4 mg/L), As (1.5 mg/L), Mn (8.4 mg/L), and Tl (up to 3.3 μg/L) in water of the Bolshoi Mukulan Brook flowing through the mines and three brooks flowing out from the base of the embankment of the tailing store no. 1. They are the major pollutants for water of the Baksan River. Upon flowing out to the plain, water of the Baksan River shows significant excess of the TLVs (in summer) for Al, Fe, Mn, Be, Si, Ti, Tl, and Hg.

  6. Ius Chasma Tributary Valleys and Adjacent Plains

    NASA Technical Reports Server (NTRS)

    2006-01-01

    This image covers valley tributaries of Ius Chasma, as well as the plains adjacent to the valleys. Ius Chasma is one of several canyons that make up the Valles Marineris canyon system. Valles Marineris likely formed by extension associated with the growth of the large volcanoes and topographic high of Tharsis to the northwest. As the ground was pulled apart, large and deep gaps resulted in the valleys seen in the top and bottom of this HiRISE image. Ice that was once in the ground could have also melted to create additional removal of material in the formation of the valleys. HiRISE is able to see the rocks along the walls of both these valleys and also impact craters in the image. Rock layers that appear lower down in elevation appear rougher and are shedding boulders. Near the top of the walls and also seen in patches along the smooth plains are brighter layers. These brighter layers are not shedding boulders so they must represent a different kind of rock formed in a different kind of environment than those further down the walls. Because they are highest in elevation, the bright layers are youngest in age. HiRISE is able to see dozens of the bright layers, which are perhaps only a meter in thickness. Darker sand dunes and ripples cover most of the plains and fill the floors of impact craters.

    Image PSP_001351_1715 was taken by the High Resolution Imaging Science Experiment (HiRISE) camera onboard the Mars Reconnaissance Orbiter spacecraft on November 9, 2006. The complete image is centered at -8.3 degrees latitude, 275.4 degrees East longitude. The range to the target site was 254.3 km (158.9 miles). At this distance the image scale ranges from 25.4 cm/pixel (with 1 x 1 binning) to 101.8 cm/pixel (with 4 x 4 binning). The image shown here has been map-projected to 25 cm/pixel and north is up. The image was taken at a local Mars time of 3:32 PM and the scene is illuminated from the west with a solar incidence angle of 59 degrees, thus the sun was about

  7. Fuel cell having electrolyte

    DOEpatents

    Wright, Maynard K.

    1989-01-01

    A fuel cell having an electrolyte control volume includes a pair of porous opposed electrodes. A maxtrix is positioned between the pair of electrodes for containing an electrolyte. A first layer of backing paper is positioned adjacent to one of the electrodes. A portion of the paper is substantially previous to the acceptance of the electrolyte so as to absorb electrolyte when there is an excess in the matrix and to desorb electrolyte when there is a shortage in the matrix. A second layer of backing paper is positioned adjacent to the first layer of paper and is substantially impervious to the acceptance of electrolyte.

  8. Mirrored serpentine flow channels for fuel cell

    DOEpatents

    Rock, Jeffrey Allan

    2000-08-08

    A PEM fuel cell having serpentine flow field channels wherein the input/inlet legs of each channel border the input/inlet legs of the next adjacent channels in the same flow field, and the output/exit legs of each channel border the output/exit legs of the next adjacent channels in the same flow field. The serpentine fuel flow channels may be longer, and may contain more medial legs, than the serpentine oxidant flow channels.

  9. View of north side from exterior stairs of adjacent building, ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    View of north side from exterior stairs of adjacent building, bottom cut off by fringed buildings, view facing south-southwest - U.S. Naval Base, Pearl Harbor, Industrial X-Ray Building, Off Sixth Street, adjacent to and south of Facility No. 11, Pearl City, Honolulu County, HI

  10. Learning Non-Adjacent Regularities at Age 0 ; 7

    ERIC Educational Resources Information Center

    Gervain, Judit; Werker, Janet F.

    2013-01-01

    One important mechanism suggested to underlie the acquisition of grammar is rule learning. Indeed, infants aged 0 ; 7 are able to learn rules based on simple identity relations (adjacent repetitions, ABB: "wo fe fe" and non-adjacent repetitions, ABA: "wo fe wo", respectively; Marcus et al., 1999). One unexplored issue is…

  11. Delayed Acquisition of Non-Adjacent Vocalic Distributional Regularities

    ERIC Educational Resources Information Center

    Gonzalez-Gomez, Nayeli; Nazzi, Thierry

    2016-01-01

    The ability to compute non-adjacent regularities is key in the acquisition of a new language. In the domain of phonology/phonotactics, sensitivity to non-adjacent regularities between consonants has been found to appear between 7 and 10 months. The present study focuses on the emergence of a posterior-anterior (PA) bias, a regularity involving two…

  12. Geomechanical analysis of a welding salt layer and its effects on adjacent sediments

    NASA Astrophysics Data System (ADS)

    Heidari, Mahdi; Nikolinakou, Maria A.; Hudec, Michael R.; Flemings, Peter B.

    2016-06-01

    We simulate welding of the source layer of a salt diapir with a forward finite-element model and study stresses and deformation in the salt layer and the diapir, as well as in their adjacent sediments. Welded salt layers are abundant in mature salt basins, where most or all of the salt has withdrawn into diapirs. However, there is little understanding of the stress field in these layers and their adjacent sediments. We show that salt flow along the source layer leads to significant stress anomalies inside the layer and in adjacent sediments. In the source layer, salt pressure becomes higher than overburden stress in nearly welded areas and becomes lower than overburden stress in adjacent thicker areas. When the source layer welds, stresses increase significantly in sediments near the weld tip, which helps compaction of these sediments and possibly their fracturing and faulting. Our model illustrates that all sediments overlying the weld experience this stress increase and the associated material changes as the weld tip propagates along the weld. We present natural examples fitting our predictions and discuss the importance of our results for the exploration, characterization, and production of reservoirs near welded salt layers.

  13. Embryonic expression of endogenous retroviral RNAs in somatic tissues adjacent to the Oikopleura germline

    PubMed Central

    Henriet, Simon; Sumic, Sara; Doufoundou-Guilengui, Carlette; Jensen, Marit Flo; Grandmougin, Camille; Fal, Kateryna; Thompson, Eric; Volff, Jean-Nicolas; Chourrout, Daniel

    2015-01-01

    Selective pressure to maintain small genome size implies control of transposable elements, and most old classes of retrotransposons are indeed absent from the very compact genome of the tunicate Oikopleura dioica. Nonetheless, two families of retrotransposons are present, including the Tor elements. The gene organization within Tor elements is similar to that of LTR retrotransposons and retroviruses. In addition to gag and pol, many Tor elements carry a third gene encoding viral envelope-like proteins (Env) that may mediate infection. We show that the Tor family contains distinct classes of elements. In some classes, env mRNA is transcribed from the 5′LTR as in retroviruses. In others, env is transcribed from an additional promoter located downstream of the 5′LTR. Tor Env proteins are membrane-associated glycoproteins which exhibit some features of viral membrane fusion proteins. Whereas some elements are expressed in the adult testis, many others are specifically expressed in embryonic somatic cells adjacent to primordial germ cells. Such embryonic expression depends on determinants present in the Tor elements and not on their surrounding genomic environment. Our study shows that unusual modes of transcription and expression close to the germline may contribute to the proliferation of Tor elements. PMID:25779047

  14. Fuel Cell Handbook update

    SciTech Connect

    Owens, W.R.; Hirschenhofer, J.H.; Engleman, R.R. Jr.; Stauffer, D.B.

    1993-11-01

    The objective of this work was to update the 1988 version of DOE`s Fuel Cell Handbook. Significant developments in the various fuel cell technologies required revisions to reflect state-of-the-art configurations and performance. The theoretical presentation was refined in order to make the handbook more useful to both the casual reader and fuel cell or systems analyst. In order to further emphasize the practical application of fuel cell technologies, the system integration information was expanded. In addition, practical elements, such as suggestions and guidelines to approximate fuel cell performance, were provided.

  15. Fuel Cell Handbook update

    NASA Astrophysics Data System (ADS)

    Owens, W. R.; Hirschenhofer, J. H.; Engleman, R. R., Jr.; Stauffer, D. B.

    The objective of this work was to update the 1988 version of DOE's Fuel Cell Handbook. Significant developments in the various fuel cell technologies required revisions to reflect state-of-the-art configurations and performance. The theoretical presentation was refined in order to make the handbook more useful to both the casual reader and fuel cell or systems analyst. In order to further emphasize the practical application of fuel cell technologies, the system integration information was expanded. In addition, practical elements, such as suggestions and guidelines to approximate fuel cell performance, were provided.

  16. Thermoelastic response of thin metal films and their adjacent materials

    SciTech Connect

    Kang, S.; Yoon, Y.; Kim, J.; Kim, W.

    2013-01-14

    A pulsed laser beam applied to a thin metal film is capable of launching an acoustic wave due to thermal expansion. Heat transfer from the thin metal film to adjacent materials can also induce thermal expansion; thus, the properties of these adjacent materials (as well as the thin metal film) should be considered for a complete description of the thermoelastic response. Here, we show that adjacent materials with a small specific heat and large thermal expansion coefficient can generate an enhanced acoustic wave and we demonstrate a three-fold increase in the peak pressure of the generated acoustic wave on substitution of parylene for polydimethylsiloxane.

  17. Alternative aircraft fuels technology

    NASA Technical Reports Server (NTRS)

    Grobman, J.

    1976-01-01

    NASA is studying the characteristics of future aircraft fuels produced from either petroleum or nonpetroleum sources such as oil shale or coal. These future hydrocarbon based fuels may have chemical and physical properties that are different from present aviation turbine fuels. This research is aimed at determining what those characteristics may be, how present aircraft and engine components and materials would be affected by fuel specification changes, and what changes in both aircraft and engine design would be required to utilize these future fuels without sacrificing performance, reliability, or safety. This fuels technology program was organized to include both in-house and contract research on the synthesis and characterization of fuels, component evaluations of combustors, turbines, and fuel systems, and, eventually, full-scale engine demonstrations. A review of the various elements of the program and significant results obtained so far are presented.

  18. Stacked Switchable Element and Diode Combination

    DOEpatents

    Branz, H. M.; Wang, Q.

    2006-06-27

    A device (10) comprises a semiconductor diode (12) and a switchable element (14) positioned in stacked adjacent relationship so that the semiconductor diode (12) and the switchable element (14) are electrically connected in series with one another. The switchable element (14) is switchable from a low-conductance state to a high-conductance state in response to the application of a forming voltage to the switchable element (14).

  19. Stacked switchable element and diode combination

    DOEpatents

    Branz, Howard M.; Wang, Qi

    2006-06-27

    A device (10) comprises a semiconductor diode (12) and a switchable element (14) positioned in stacked adjacent relationship so that the semiconductor diode (12) and the switchable element (14) are electrically connected in series with one another. The switchable element (14) is switchable from a low-conductance state to a high-conductance state in response to the application of a forming voltage to the switchable element (14).

  20. Impaired statistical learning of non-adjacent dependencies in adolescents with specific language impairment

    PubMed Central

    Hsu, Hsinjen J.; Tomblin, J. Bruce; Christiansen, Morten H.

    2014-01-01

    Being able to track dependencies between syntactic elements separated by other constituents is crucial for language acquisition and processing (e.g., in subject-noun/verb agreement). Although long assumed to require language-specific machinery, research on statistical learning has suggested that domain-general mechanisms may support the acquisition of non-adjacent dependencies. In this study, we investigated whether individuals with specific language impairment (SLI)—who have problems with long-distance dependencies in language—also have problems with statistical learning of non-adjacent relations. The results confirmed this hypothesis, indicating that statistical learning may subserve the acquisition and processing of long-distance dependencies in natural language. PMID:24639661

  1. Postirradiation examination of HTR fuel

    SciTech Connect

    Nabielek, H.; Reitsamer, G.; Kania, M.J.

    1986-01-01

    Fuel for the High Temperature Reactor (HTR) consists of 1 mm diameter coated particles uniformly distributed in a graphite matrix within a cold-molded 60 mm diameter spherical fuel element. Fuel performance demonstrations under simulated normal operation conditions are conducted in accelerated neutron environments available in Material Test Reactors and in real-time environments such as the Arbeitsgemeinschaft Versuchsreaktor (AVR) Juelich. Postirradiation examinations are then used to assess fuel element behavior and the detailed performance of the coated particles. The emphasis in postirradiation examination and accident testing is on assessment of the capability for fuel elements and individual coated particles to retain fission products and actinide fuel materials. To accomplish this task, techniques have been developed which measures fission product and fuel material distributions within or exterior to the particle: Hot Gas Chlorination - provides an accurate method to measure total fuel material concentration outside intact particles; Profile Electrolytic Deconsolidation - permits determination of fission product distribution along fuel element diameter and retrieval of fuel particles from positions within element; Gamma Spectrometry - provides nondestructive method to measure defect particle fractions based on retention of volatile metallic fission products; Particle Cracking - permits a measure of the partitioning of fission products between fuel kernel and particle coatings, and the derivation of diffusion parameters in fuel materials; Micro Gas Analysis - provides gaseous fission product and reactive gas inventory within free volume of single particles; and Mass-spectrometric Burnup Determination - utilizes isotope dilution for the measurement of heavy metal isotope abundances.

  2. 73. PASSAGE ADJACENT TO ROOM 232, EAST WING, SECOND FLOOR, ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    73. PASSAGE ADJACENT TO ROOM 232, EAST WING, SECOND FLOOR, LOOKING WEST BY NORTHWEST, SHOWING EASTERNMOST ARCH OF FORMER GREAT HALL NORTH ARCADE - Smithsonian Institution Building, 1000 Jefferson Drive, between Ninth & Twelfth Streets, Southwest, Washington, District of Columbia, DC

  3. View of viaduct, looking SE from roof of adjacent parking ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    View of viaduct, looking SE from roof of adjacent parking garage. - Mulberry Street Viaduct, Spanning Paxton Creek & Cameron Street (State Route 230) at Mulberry Street (State Route 3012), Harrisburg, Dauphin County, PA

  4. Cement Leakage into Adjacent Vertebral Body Following Percutaneous Vertebroplasty

    PubMed Central

    Park, Jae Hoo; Kim, Hyeun Sung

    2016-01-01

    Percutaneous vertebroplasty (PV) is a minimally invasive procedure for osteoporotic vertebral compression fractures that fail to respond to conventional conservative treatment. It significantly improves intolerable back pain within hours, and has a low complication rate. Although rare, PV is not free of complications, most of which are directly related to cement leakage. Because of its association with new adjacent fracture, the importance of cement leakage into the adjacent disc space is paramount. Here, we report an interesting case of cement leakage into the adjacent upper vertebral body as well as disc space following PV. To the best of our knowledge, there has been no report of cement leakage into the adjacent vertebral body following PV. This rare case is presented along with a review of the literature. PMID:27437018

  5. 1. HEBRONVILLE MILL COMPLEX ADJACENT TO NORTHEAST CORRIDOR. HEBRONVILLE, BRISTOL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    1. HEBRONVILLE MILL COMPLEX ADJACENT TO NORTHEAST CORRIDOR. HEBRONVILLE, BRISTOL CO., MA. Sec. 4116, MP 193.75. - Northeast Railroad Corridor, Amtrak Route between RI/MA State Line & South Station, Boston, Suffolk County, MA

  6. 3. DODGEVILLE MILL COMPLEX ADJACENT TO NORTHEAST CORRIDOR DODGEVILLE, BRISTOL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    3. DODGEVILLE MILL COMPLEX ADJACENT TO NORTHEAST CORRIDOR DODGEVILLE, BRISTOL CO., MA. Sec. 4116, MP 195.55. - Northeast Railroad Corridor, Amtrak Route between RI/MA State Line & South Station, Boston, Suffolk County, MA

  7. 33. HISTORIC PLAQUE MARKING WHERE JOHNSTON DIED, ADJACENT TO PATHWAY ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    33. HISTORIC PLAQUE MARKING WHERE JOHNSTON DIED, ADJACENT TO PATHWAY WITH CONCRETE CULVERT LEADING NORTH OUT OF RAVINE TOWARD JOHNSTON MEMORIAL SITE. VIEW NW. - Shiloh National Military Park Tour Roads, Shiloh, Hardin County, TN

  8. Lock 4 View east of lock wall and adjacent ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Lock 4 - View east of lock wall and adjacent roadway built atop tow path. The gate pocket can be seen at center. - Savannah & Ogeechee Barge Canal, Between Ogeechee & Savannah Rivers, Savannah, Chatham County, GA

  9. 1. Ninth Street (west) facade. Adjacent on the north is ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    1. Ninth Street (west) facade. Adjacent on the north is the 9th Street facade of 816 E Street. Both buildings were originally one property. - Riley Building, Rendezvous Adult Magazines & Films, 437 Ninth Street, Northwest, Washington, District of Columbia, DC

  10. 2. THREEQUARTER VIEW FROM ADJACENT ACCESS ROAD SHOWING THREE SPANS ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    2. THREE-QUARTER VIEW FROM ADJACENT ACCESS ROAD SHOWING THREE SPANS AND NORTHWEST APPROACH SPANS, LOOKING SOUTHEAST - Red River Bridge, Spanning Red River at U.S. Highway 82, Garland, Miller County, AR

  11. 1. VIEW FROM ROOFTOP OF BUILDING (MOTEL) ADJACENT TO TECHWOOD ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    1. VIEW FROM ROOFTOP OF BUILDING (MOTEL) ADJACENT TO TECHWOOD HOMES, LOOKING SOUTH. GARAGE TO EXTREME LEFT, BUILDING 1 TO EXTREME RIGHT. - Techwood Homes (Public Housing), Bounded by North Avenue, Parker Street, William Street & Lovejoy Street, Atlanta, Fulton County, GA

  12. 3. View of north side of house facing from adjacent ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    3. View of north side of house facing from adjacent vacant property. Original wood lap siding and trim is covered by aluminum siding. Recessed side porch is in middle. - 645 South Eighteenth Street (House), Louisville, Jefferson County, KY

  13. 1. A BRICK AND CONCRETE FAN HOUSING ADJACENT TO ONE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    1. A BRICK AND CONCRETE FAN HOUSING ADJACENT TO ONE OF THE ADIT OPENINGS (VIEW TO THE NORTH). - Foster Gulch Mine, Fan Housing, Bear Creek 1 mile Southwest of Town of Bear Creek, Red Lodge, Carbon County, MT

  14. 7. August, 1970 9 ORANGE STREET, ADJACENT TO UNITARIAN CHURCH ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    7. August, 1970 9 ORANGE STREET, ADJACENT TO UNITARIAN CHURCH (NOT IN STUDY AREA) - Orange & Union Streets Neighborhood Study, 8-31 Orange Street, 9-21 Union Street & Stone Alley, Nantucket, Nantucket County, MA

  15. OBLIQUE OF SOUTHWEST END AND SOUTHEAST SIDE, WITH ADJACENT FACILITY ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    OBLIQUE OF SOUTHWEST END AND SOUTHEAST SIDE, WITH ADJACENT FACILITY 391 IN THE FOREGROUND. - U.S. Naval Base, Pearl Harbor, Joint Intelligence Center, Makalapa Drive in Makalapa Administration Area, Pearl City, Honolulu County, HI

  16. Complications in exodontia--accidental dislodgment to adjacent anatomical areas.

    PubMed

    Grandini, S A; Barros, V M; Salata, L A; Rosa, A L; Soares, U N

    1993-01-01

    The authors report 4 cases of accidental dislodgement of teeth to adjacent anatomical areas during extraction. The causes and their prevention are discussed and solutions for the problem are suggested. PMID:8241759

  17. 6. Detail, vertical guides adjacent to east portal of Tunnel ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    6. Detail, vertical guides adjacent to east portal of Tunnel 28, view to southwest, 135mm lens with electronic flash fill. - Central Pacific Transcontinental Railroad, Tunnel No. 28, Milepost 134.75, Applegate, Placer County, CA

  18. VIEW OF CONSTRUCTION CAMP ROCK FEATURE WITH OVER, ADJACENT TO ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    VIEW OF CONSTRUCTION CAMP ROCK FEATURE WITH OVER, ADJACENT TO THE COLUMBIA SOUTHERN CANAL. LOOKING NORTHWEST - Tumalo Irrigation District, Tumalo Project, West of Deschutes River, Tumalo, Deschutes County, OR

  19. Pump house adjacent to the superintendent's house at the west ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Pump house adjacent to the superintendent's house at the west end of the complex near Highway 101. Detail of Holloshaft pump. View to the south. - Prairie Creek Fish Hatchery, Hwy. 101, Orick, Humboldt County, CA

  20. VIEW OF NORTHERN AND EASTERN SIDES FROM PARKING LOT ADJACENT ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    VIEW OF NORTHERN AND EASTERN SIDES FROM PARKING LOT ADJACENT TO BUILDING 199 (POLICE STATION) - U.S. Naval Base, Pearl Harbor, Post Office, Avenue A near Eleventh Avenue, Pearl City, Honolulu County, HI

  1. 24. INTERIOR VIEW, WILLIAM GRAY AT SIZING GUAGE ADJACENT TO ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    24. INTERIOR VIEW, WILLIAM GRAY AT SIZING GUAGE ADJACENT TO BRADLEY HAMMER; NOTE THIS IS THE SAME TOOL AS BEING FORGED ABOVE - Warwood Tool Company, Foot of Nineteenth Street, Wheeling, Ohio County, WV

  2. Detail exterior view looking north showing piping system adjacent to ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Detail exterior view looking north showing piping system adjacent to engine house. Gas cooling system is on far right. - Burnsville Natural Gas Pumping Station, Saratoga Avenue between Little Kanawha River & C&O Railroad line, Burnsville, Braxton County, WV

  3. VIEW OF LAMP FIXTURE (EXTERIOR) ADJACENT TO ENTRANCE AT SOUTHWEST ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    VIEW OF LAMP FIXTURE (EXTERIOR) ADJACENT TO ENTRANCE AT SOUTHWEST CORNER OF BUILDING 23, FACING NORTH - Roosevelt Base, Auditorium-Gymnasium, West Virginia Street between Richardson & Reeves Avenues, Long Beach, Los Angeles County, CA

  4. 14. Charles Acey Cobb standing adjacent to the fish screen ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    14. Charles Acey Cobb standing adjacent to the fish screen he designed and installed in the Congdon Canal, facing southeast. Photo dates ca. late 1920's. - Congdon Canal, Fish Screen, Naches River, Yakima, Yakima County, WA

  5. 52. EASTSIDE PLANT: GENERAL VIEW OF GOVERNOR ADJACENT TO GENERATOR ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    52. EASTSIDE PLANT: GENERAL VIEW OF GOVERNOR ADJACENT TO GENERATOR - American Falls Water, Power & Light Company, Island Power Plant, Snake River, below American Falls Dam, American Falls, Power County, ID

  6. Interior building details of Building A, dungeon cell adjacent to ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Interior building details of Building A, dungeon cell adjacent to northwest cell: granite and brick threshold, poured concrete floors, plastered finished walls, vaulted veiling; northwesterly view - San Quentin State Prison, Building 22, Point San Quentin, San Quentin, Marin County, CA

  7. VIEW OF CONCRETE CHANNEL ADJACENT TO TUMALO FEED CANAL INTAKE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    VIEW OF CONCRETE CHANNEL ADJACENT TO TUMALO FEED CANAL INTAKE STRUCTURE (DOWNSTREAM SIDE). LOOKING EAST/NORTHEAST - Tumalo Irrigation District, Tumalo Project, West of Deschutes River, Tumalo, Deschutes County, OR

  8. Hydrologic Connection Between Geysers and Adjacent Thermal Pools, Two Examples: El Tatio, Chile and Yellowstone, USA

    NASA Astrophysics Data System (ADS)

    Munoz Saez, C.; Fauria, K.; Manga, M.; Hurwitz, S.; Namiki, A.

    2014-12-01

    Geyser eruption cycles can be influenced by adjacent and distant thermals sources, suggesting a hydraulic connection through permeable pathways. Diffusion of fluid pressure can be responsible for the communication between geysers. In this study we examine the processes linking two different geysers with adjacent thermal pools. The first was Vega Rinconada, located at El Tatio geyser field, Chile, where we measured temperature inside the conduit between the ground surface and a depth of seven meters, at one-meter intervals. The second was Lone Star Geyser in Yellowstone National Park, where we measured temperature of the overflow water at the base of the cone. Concurrently, we measured temperature and the water level in pools adjacent to both geysers. We found common elements in both geyser - pool systems: First, water temperature in both adjacent pools was below the boiling point and cooler than water in the geysers. Second, changes in pool water levels were correlated with eruptions of the geysers. During the quiescent period of the geysers, the water level increased in adjacent pools, while water level in the pools deceased during eruptions. Additionally, measurements inside of the conduit in Vega Rinconada Geyser showed that water temperature increased in the deepest part of the conduit during eruptions, while water temperature decreased in the shallow part of the geyser conduit (~1 to 2 m). These drops in temperature in the shallow conduit were coincident with the drop in water level in the adjacent pool. This suggests that after the initiation of an eruption, water may drain from the pool to the geyser. Furthermore, we observed a temperature drop of 3oC in the shallow conduit immediately preceding the end of an eruption. This suggests that flow from the pool to geyser contributes to eruption shut off. Our observations of geyser-pool systems indicate a hydrologic connection between the geysers and their adjacent pools. In the case of Vega Rinconada, cold water

  9. Adjacent Segment Disease Perspective and Review of the Literature

    PubMed Central

    Saavedra-Pozo, Fanor M.; Deusdara, Renato A. M.; Benzel, Edward C.

    2014-01-01

    Background Adjacent segment disease has become a common topic in spine surgery circles because of the significant increase in fusion surgery in recent years and the development of motion preservation technologies that theoretically should lead to a decrease in this pathology. The purpose of this review is to organize the evidence available in the current literature on this subject. Methods For this literature review, a search was conducted in PubMed with the following keywords: adjacent segment degeneration and disease. Selection, review, and analysis of the literature were completed according to level of evidence. Results The PubMed search identified 850 articles, from which 41 articles were selected and reviewed. The incidence of adjacent segment disease in the cervical spine is close to 3% without a significant statistical difference between surgical techniques (fusion vs arthroplasty). Authors report the incidence of adjacent segment disease in the lumbar spine to range from 2% to 14%. Damage to the posterior ligamentous complex and sagittal imbalances are important risk factors for both degeneration and disease. Conclusion Insufficient evidence exists at this point to support the idea that total disc arthroplasty is superior to fusion procedures in minimizing the incidence of adjacent segment disease. The etiology is most likely multifactorial but it is becoming abundantly clear that adjacent segment disease is not caused by motion segment fusion alone. Fusion plus the presence of abnormal end-fusion alignment appears to be a major factor in creating end-fusion stresses that result in adjacent segment degeneration and subsequent disease. The data presented cast further doubt on previously established rationales for total disc arthroplasty, at least with regard to the effect of total disc arthroplasty on adjacent segment degeneration pathology. PMID:24688337

  10. Approximating the largest eigenvalue of network adjacency matrices

    NASA Astrophysics Data System (ADS)

    Restrepo, Juan G.; Ott, Edward; Hunt, Brian R.

    2007-11-01

    The largest eigenvalue of the adjacency matrix of a network plays an important role in several network processes (e.g., synchronization of oscillators, percolation on directed networks, and linear stability of equilibria of network coupled systems). In this paper we develop approximations to the largest eigenvalue of adjacency matrices and discuss the relationships between these approximations. Numerical experiments on simulated networks are used to test our results.

  11. Fuel cell generator energy dissipator

    SciTech Connect

    Veyo, S.E.; Dederer, J.T.; Gordon, J.T.; Shockling, L.A.

    2000-02-15

    An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a fuel cell generator when the electrical power output of the fuel cell generator is terminated. During a generator shut down condition, electrically resistive elements are automatically connected across the fuel cell generator terminals in order to draw current, thereby depleting the fuel inventory in the generator. The invention provides a safety function in eliminating the fuel energy, and also provides protection to the fuel cell stack by eliminating overheating.

  12. NACA Research on Slurry Fuels

    NASA Technical Reports Server (NTRS)

    Pinns, M L; Olson, W T; Barnett, H C; Breitwieser, R

    1958-01-01

    An extensive program was conducted to investigate the use of concentrated slurries of boron and magnesium in liquid hydrocarbon as fuels for afterburners and ramjet engines. Analytical calculations indicated that magnesium fuel would give greater thrust and that boron fuel would give greater range than are obtainable from jet hydrocarbon fuel alone. It was hoped that the use of these solid elements in slurry form would permit the improvement to be obtained without requiring unconventional fuel systems or combustors. Small ramjet vehicles fueled with magnesium slurry were flown successfully, but the test flights indicated that further improvement of combustors and fuel systems was needed.

  13. Storage of spent fuel from the nation`s nuclear reactors: Status, technology, and policy options

    SciTech Connect

    Not Available

    1989-10-01

    Since the beginning of the commercial nuclear electric power industry, it has been recognized that spent nuclear reactor fuel must be able to be readily removed from the reactor vessel in the plant and safely stored on-site. The need for adjacent ready storage is first for safety. In the event of an emergency, or necessary maintenance that requires the removal of irradiated fuel from the reactor vessel, cooled reserve storage capacity for the full amount of fuel from the reactor core must be available. Also, the uranium fuel in the reactor eventually reaches the point where its heat generation is below the planned efficiency for steam production which drives the turbines and generators. It then must be replaced by fresh uranium fuel, with the ``spent fuel`` elements being removed to a safe and convenient storage location near the reactor vessel. The federal nuclear waste repository program, even without delays in the current schedule of disposal becoming available in 2003, will result in a large percentage of the 111 existing operable commercial reactors requiring expansion of their spent fuel storage capacity. How that need can and will be met raises issues of both technology and policy that will be reviewed in this report.

  14. Storage of spent fuel from the nation's nuclear reactors: Status, technology, and policy options

    SciTech Connect

    Not Available

    1989-10-01

    Since the beginning of the commercial nuclear electric power industry, it has been recognized that spent nuclear reactor fuel must be able to be readily removed from the reactor vessel in the plant and safely stored on-site. The need for adjacent ready storage is first for safety. In the event of an emergency, or necessary maintenance that requires the removal of irradiated fuel from the reactor vessel, cooled reserve storage capacity for the full amount of fuel from the reactor core must be available. Also, the uranium fuel in the reactor eventually reaches the point where its heat generation is below the planned efficiency for steam production which drives the turbines and generators. It then must be replaced by fresh uranium fuel, with the spent fuel'' elements being removed to a safe and convenient storage location near the reactor vessel. The federal nuclear waste repository program, even without delays in the current schedule of disposal becoming available in 2003, will result in a large percentage of the 111 existing operable commercial reactors requiring expansion of their spent fuel storage capacity. How that need can and will be met raises issues of both technology and policy that will be reviewed in this report.

  15. 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.

  16. The Transuranium Elements.

    ERIC Educational Resources Information Center

    Seaborg, Glenn T.

    1985-01-01

    Discusses the unusual chemistry of the transuranium elements as well as their impact on the periodic table. Also considers the practical applications of transuranium isotopes, such as their use in nuclear fuel for the large-scale generation of electricity. (JN)

  17. NUCLEAR REACTOR ELEMENT

    DOEpatents

    Sanz, M.C.; Scully, C.N.

    1961-06-27

    The patented fuel element is a hexagonal graphite body having an axial channel therethrough. The graphite is impregnated with uranium which is concentrated near the axial channel. Layers of tantalum nitride and tantalum carbide are disposed on the surface of the body confronting the channel.

  18. Corrosion free phosphoric acid fuel cell

    DOEpatents

    Wright, Maynard K.

    1990-01-01

    A phosphoric acid fuel cell with an electrolyte fuel system which supplies electrolyte via a wick disposed adjacent a cathode to an absorbent matrix which transports the electrolyte to portions of the cathode and an anode which overlaps the cathode on all sides to prevent corrosion within the cell.

  19. Nuclear core and fuel assemblies

    DOEpatents

    Downs, Robert E.

    1981-01-01

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

  20. 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. Fuel system bubble dissipation device

    SciTech Connect

    Iseman, W.J.

    1987-11-03

    This patent describes a bubble dissipation device for a fuel system wherein fuel is delivered through a fuel line from a fuel tank to a fuel control with the pressure of the fuel being progressively increased by components including at least one pump stage and an ejector in advance of the pump state. The ejector an ejector casing with a wall defining an elongate tubular flow passage which forms a portion of the fuel line to have all of the fuel flow through the tubular flow passage in flowing from the fuel tank to the fuel control, a nozzle positioned entirely within the tubular flow passage and spaced from the wall to permit fuel flow. The nozzle has an inlet and an outlet with the inlet connected to the pump stage to receive fuel under pressure continuously from the pump stage, a bubble accumulation chamber adjoining and at a level above the ejector casing and operatively connected to the fuel line in advance of the ejector casing. The bubble accumulation chamber is of a size to function as a fuel reservoir and hold an air bubble containing vapor above the level of fuel therein and having an outlet adjacent the bottom thereof operatively connected to the tubular flow passage in the ejector casing at an inlet end, a bubble accumulation chamber inlet above the level of the bubble accumulation chamber outlet whereby fuel can flow through the bubble accumulation chamber from the inlet to the outlet thereof with a bubble in the fuel rising above the fuel level in the bubble accumulation chamber.

  2. Cations Stiffen Actin Filaments by Adhering a Key Structural Element to Adjacent Subunits.

    PubMed

    Hocky, Glen M; Baker, Joseph L; Bradley, Michael J; Sinitskiy, Anton V; De La Cruz, Enrique M; Voth, Gregory A

    2016-05-26

    Ions regulate the assembly and mechanical properties of actin filaments. Recent work using structural bioinformatics and site-specific mutagenesis favors the existence of two discrete and specific divalent cation binding sites on actin filaments, positioned in the long axis between actin subunits. Cation binding at one site drives polymerization, while the other modulates filament stiffness and plays a role in filament severing by the regulatory protein, cofilin. Existing structural methods have not been able to resolve filament-associated cations, and so in this work we turn to molecular dynamics simulations to suggest a candidate binding pocket geometry for each site and to elucidate the mechanism by which occupancy of the "stiffness site" affects filament mechanical properties. Incorporating a magnesium ion in the "polymerization site" does not seem to require any large-scale change to an actin subunit's conformation. Binding of a magnesium ion in the "stiffness site" adheres the actin DNase-binding loop (D-loop) to its long-axis neighbor, which increases the filament torsional stiffness and bending persistence length. Our analysis shows that bound D-loops occupy a smaller region of accessible conformational space. Cation occupancy buries key conserved residues of the D-loop, restricting accessibility to regulatory proteins and enzymes that target these amino acids. PMID:27146246

  3. Characteristics of radiation porosity formed upon irradiation in a BN-600 reactor in the fuel-element cans of cold-deformed steel EK-164 (06Kh16N20M2G2BTFR)-ID c.d.

    NASA Astrophysics Data System (ADS)

    Portnykh, I. A.; Kozlov, A. V.; Panchenko, V. L.; Mitrofanova, N. M.

    2012-05-01

    At present, it is the austenitic cold-deformed steel EK164 (06Kh16N20M2G2BTFR)-ID that is considered as a promising material for the achievement of a maximum damage (no less than 110 dpa) and maximum burnup (≥15%). In this work, we have determined the characteristics of porosity formed upon irradiation in a BN-600 reactor to the maximum damaging dose of 77 dpa in the materials of fuel-element cans made of cold-deformed steel EK164-ID c.d. A comparison has been made with analogous characteristics obtained earlier using the standard material, i.e., the cold-deformed steel ChS68 (06Kh16N 15M2G2TFR)-ID c.d.

  4. Eddy current measurement of tube element spacing

    DOEpatents

    Latham, Wayne Meredith; Hancock, Jimmy Wade; Grut, Jayne Marie

    1998-01-01

    A method of electromagnetically measuring the distance between adjacent tube elements in a heat exchanger. A cylindrical, high magnetic permeability ferrite slug is placed in the tube adjacent the spacing to be measured. A bobbin or annular coil type probe operated in the absolute mode is inserted into a second tube adjacent the spacing to be measured. From prior calibrations on the response of the eddy current coil, the signals from the coil, when sensing the presence of the ferrite slug, are used to determine the spacing between the tubes.

  5. Laplacian versus adjacency matrix in quantum walk search

    NASA Astrophysics Data System (ADS)

    Wong, Thomas G.; Tarrataca, Luís; Nahimov, Nikolay

    2016-06-01

    A quantum particle evolving by Schrödinger's equation contains, from the kinetic energy of the particle, a term in its Hamiltonian proportional to Laplace's operator. In discrete space, this is replaced by the discrete or graph Laplacian, which gives rise to a continuous-time quantum walk. Besides this natural definition, some quantum walk algorithms instead use the adjacency matrix to effect the walk. While this is equivalent to the Laplacian for regular graphs, it is different for non-regular graphs and is thus an inequivalent quantum walk. We algorithmically explore this distinction by analyzing search on the complete bipartite graph with multiple marked vertices, using both the Laplacian and adjacency matrix. The two walks differ qualitatively and quantitatively in their required jumping rate, runtime, sampling of marked vertices, and in what constitutes a natural initial state. Thus the choice of the Laplacian or adjacency matrix to effect the walk has important algorithmic consequences.

  6. FUEL ASSAY REACTOR

    DOEpatents

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  7. Reprocessing RERTR silicide fuels

    SciTech Connect

    Rodrigues, G.C.; Gouge, A.P.

    1983-05-01

    The Reduced Enrichment Research and Test Reactor Program is one element of the United States Government's nonproliferation effort. High-density, low-enrichment, aluminum-clad uranium silicide fuels may be substituted for the highly enriched aluminum-clad alloy fuels now in use. Savannah River Laboratory has performed studies which demonstrate reprocessability of spent RERTR silicide fuels at Savannah River Plant. Results of dissolution and feed preparation tests and solvent extraction processing demonstrations with both unirradiated and irradiated uranium silicide fuels are presented.

  8. Reprocessing RERTR fuels

    SciTech Connect

    Rodrigues, G.C.

    1983-01-01

    The Reduced Enrichment Research and Test Reactor Program is one element of the United States Government's nonproliferation effort. High density, low enrichment aluminum-clad dispersed uranium compound fuels may be substituted for the highly enriched aluminum-clad aluminum-uranium alloy fuels now in use. Savannah River Laboratory has performed studies which demonstrate reprocessability of spent RERTR fuels at Savannah River Plant. Results of dissolution and feed preparation tests with both unirradiated and irradiated (up to approximately 90% burnup) fuels are presented. 13 references, 2 figures, 4 tables.

  9. On the Adjacent Eccentric Distance Sum Index of Graphs

    PubMed Central

    Qu, Hui; Cao, Shujuan

    2015-01-01

    For a given graph G, ε(v) and deg(v) denote the eccentricity and the degree of the vertex v in G, respectively. The adjacent eccentric distance sum index of a graph G is defined as ξsv(G)=∑v∈V(G)ε(v)D(v)deg(v), where D(v)=∑u∈V(G)d(u,v) is the sum of all distances from the vertex v. In this paper we derive some bounds for the adjacent eccentric distance sum index in terms of some graph parameters, such as independence number, covering number, vertex connectivity, chromatic number, diameter and some other graph topological indices. PMID:26091095

  10. Nonlinear spin wave coupling in adjacent magnonic crystals

    NASA Astrophysics Data System (ADS)

    Sadovnikov, A. V.; Beginin, E. N.; Morozova, M. A.; Sharaevskii, Yu. P.; Grishin, S. V.; Sheshukova, S. E.; Nikitov, S. A.

    2016-07-01

    We have experimentally studied the coupling of spin waves in the adjacent magnonic crystals. Space- and time-resolved Brillouin light-scattering spectroscopy is used to demonstrate the frequency and intensity dependent spin-wave energy exchange between the side-coupled magnonic crystals. The experiments and the numerical simulation of spin wave propagation in the coupled periodic structures show that the nonlinear phase shift of spin wave in the adjacent magnonic crystals leads to the nonlinear switching regime at the frequencies near the forbidden magnonic gap. The proposed side-coupled magnonic crystals represent a significant advance towards the all-magnonic signal processing in the integrated magnonic circuits.

  11. Conceptual Design of a Clinical BNCT Beam in an Adjacent Dry Cell of the Jozef Stefan Institute TRIGA Reactor

    SciTech Connect

    Maucec, Marko

    2000-11-15

    The MCNP4B Monte Carlo transport code is used in a feasibility study of the epithermal neutron boron neutron capture therapy facility in the thermalizing column of the 250-kW TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). To boost the epithermal neutron flux at the reference irradiation point, the efficiency of a fission plate with almost 1.5 kg of 20% enriched uranium and 2.3 kW of thermal power is investigated. With the same purpose in mind, the TRIGA reactor core setup is optimized, and standard fresh fuel elements are concentrated partly in the outermost ring of the core. Further, a detailed parametric study of the materials and dimensions for all the relevant parts of the irradiation facility is carried out. Some of the standard epithermal neutron filter/moderator materials, as well as 'pressed-only' low-density Al{sub 2}O{sub 3} and AlF{sub 3}, are considered. The proposed version of the BNCT facility, with PbF{sub 2} as the epithermal neutron filter/moderator, provides an epithermal neutron flux of {approx}1.1 x 10{sup 9} n/cm{sup 2}.s, thus enabling patient irradiation times of <60 min. With reasonably low fast neutron and photon contamination ([overdot]D{sub nfast}/{phi}{sub epi} < 5 x 10{sup -13} Gy.cm{sup 2}/n and [overdot]D{sub {gamma}} /{phi}{sub epi} < 3 x 10{sup -13} Gy.cm{sup 2}/n), the in-air performances of the proposed beam are comparable to all existing epithermal BNCT facilities. The design presents an equally efficient alternative to the BNCT beams in TRIGA reactor thermal columns that are more commonly applied. The cavity of the dry cell, a former JSI TRIGA reactor spent-fuel storage facility, adjacent to the thermalizing column, could rather easily be rearranged into a suitable patient treatment room, which would substantially decrease the overall developmental costs.

  12. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    SciTech Connect

    Webb, B.J.

    1988-01-01

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab.

  13. The optimization of an AP1000 fuel assembly for the transmutation of plutonium and minor actinides

    NASA Astrophysics Data System (ADS)

    Washington, Jeremy A.

    The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. The goal of this thesis is to examine the potential of light water reactors for plutonium and minor actinides transmutation as a near-term solution. This thesis screens the available nuclear isotope database to identify potential absorbers as coatings on a transmutation fuel in a light water reactor. A spectral shift absorber coating tunes the neutron energy spectrum experienced by the underlying target fuel. Eleven different spectral shift absorbers (B4C, CdO, Dy2O3, Er 2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) have been selected for further evaluation. A model developed using the NEWT module of SCALE 6.1 code provided performance data for the burnup of the target fuel rods. Irradiation of the target fuels occurs in a Westinghouse 17x17 XL Robust Fuel Assembly over a 1400 Effective Full Power Days (EFPD) interval. The fuels evaluated in this thesis include PuO2, Pu3Si2, PuN, MOX, PuZrH, PuZrHTh, PuZrO 2, and PuUZrH. MOX (5 wt% PuO2), Pu0.31ZrH 1.6Th1.08, and PuZrO2MgO (8 wt%) are selected for detailed analysis in a multi-pin transmutation assembly. A coupled model optimized the resulting transmutation fuel elements. The optimization considered three stages of fuel assemblies containing target fuel pins. The first stage optimized four target fuel pins adjacent to the central instrumentation channel. The second stage evaluated a variety of assemblies with multiple target fuel pins and the third stage re-optimized target fuel pins in the second-stage assembly. A PuZrO2MgO (8 wt%) target fuel with a coating of Lu 2O3 resulted in the greatest reduction in curium-244

  14. Toxicity and bioavailability of metals in the Missouri River adjacent to a lead refinery

    USGS Publications Warehouse

    Chapman, Duane C.; Allert, Ann L.; Fairchild, James F.; May, Thomas W.; Schmitt, Christopher J.; Callahan, Edward V.

    2001-01-01

    This study is an evaluation of the potential environmental impacts of contaminated groundwater from the ASARCO metals refining facility adjacent to the Missouri River in Omaha, Nebraska. Surface waters, sediments, and sediment pore waters were collected from the Burt-Izard drain, which transects the facility, and from the Missouri River adjacent to the facility. Groundwater was also collected from the facility. Waters and sediments were analyzed for inorganic contaminants, and the toxicity of the waters was evaluated with the Ceriodaphnia dubia 7-day test. Concentrations of several elemental contaminants were highly elevated in the groundwater, but not in river sediment pore waters. Lead concentrations were moderately elevated in whole sediment at one site, but lead concentrations in pore waters were low due to apparent sequestration by acid-volatile sulfides. The groundwater sample was highly toxic to C. dubia, causing 100% mortality. Even at the lowest groundwater concentration tested (6.25%) C. dubia survival was reduced; however, at that concentration, reproduction was not significantly different from upstream porewater reference samples. Sediment pore waters were not toxic, except reproduction in pore water collected from one downstream site was somewhat reduced. The decrease in reproduction could not be attributed to measured elemental contaminants.

  15. It's elemental

    NASA Astrophysics Data System (ADS)

    The Periodic Table of the elements will now have to be updated. An international team of researchers has added element 110 to the Earth's armory of elements. Though short-lived—of the order of microseconds, element 110 bottoms out the list as the heaviest known element on the planet. Scientists at the Heavy Ion Research Center in Darmstadt, Germany, made the 110-proton element by colliding a lead isotope with nickel atoms. The element, which is yet to be named, has an atomic mass of 269.

  16. 7. VIEW OF WATER TREATMENT PLANT, ADJACENT TO THE COAL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    7. VIEW OF WATER TREATMENT PLANT, ADJACENT TO THE COAL CONVEYOR; IN THE DISTANCE IS THE FREQUENCY CHANGER HOUSE, WHICH IS ATTACHED TO SWITCH HOUSE NO. 1; LOOKING WEST. - Commonwealth Electric Company, Fisk Street Electrical Generating Station, 1111 West Cermak Avenue, Chicago, Cook County, IL

  17. Colposcopy of vaginal and vulvar human papillomavirus and adjacent sites.

    PubMed

    Hatch, K

    1993-03-01

    Human papillomaviral infections can affect the entire lower female genital tract as multifocal or multicentric disease as well as the surrounding anatomic and adjacent sites. The traditional colposcopic methods are necessary to assist in the diagnosis and help differentiate these infections from other disease mimics. PMID:8392676

  18. Biogeochemistry of hydrothermally and adjacent non-altered soils

    Technology Transfer Automated Retrieval System (TEKTRAN)

    As a field/lab project, students in the Soil Biogeochemistry class of the University of Nevada, Reno described and characterized seven pedons, developed in hydrothermally and adjacent non-hydrothermally altered andesitic parent material near Reno, NV. Hydrothermally altered soils had considerably lo...

  19. 22. Float located adjacent to entry stair in filtration bed. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    22. Float located adjacent to entry stair in filtration bed. The float actuates a valve that maintains water level over the bed. - Lake Whitney Water Filtration Plant, Filtration Plant, South side of Armory Street between Edgehill Road & Whitney Avenue, Hamden, New Haven County, CT

  20. 2. VIEW FROM ROOFTOP OF BUILDING (MOTEL) ADJACENT TO TECHWOOD ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    2. VIEW FROM ROOFTOP OF BUILDING (MOTEL) ADJACENT TO TECHWOOD HOMES, LOOKING WEST. GEORGIA TECH DORMITORY BUILDING, 581-587 TECHWOOD DRIVE, IN FOREGROUND. - Techwood Homes (Public Housing), Bounded by North Avenue, Parker Street, William Street & Lovejoy Street, Atlanta, Fulton County, GA