Science.gov

Sample records for advanced fuel elements

  1. 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.

  2. 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].

  3. 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)

  4. 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.

  5. 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)

  6. 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)

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. 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)

  13. 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.

  14. 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)

  15. 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.

  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. 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)

  20. 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.

  1. 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)

  2. 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)

  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. 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

  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. 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)

  10. 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.

  11. 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.

  12. 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)

  13. 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.

  14. 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.

  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. Polarized advanced fuel reactors

    SciTech Connect

    Kulsrud, R.M.

    1987-07-01

    The d-/sup 3/He reaction has the same spin dependence as the d-t reaction. It produces no neutrons, so that if the d-d reactivity could be reduced, it would lead to a neutron-lean reactor. The current understanding of the possible suppression of the d-d reactivity by spin polarization is discussed. The question as to whether a suppression is possible is still unresolved. Other advanced fuel reactions are briefly discussed. 11 refs.

  17. 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.

  18. 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)

  19. 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)

  20. 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)

  1. 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.

  2. 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.

  3. 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)

  4. 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.

  5. 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)

  6. Advanced Fuels Campaign 2012 Accomplishments

    SciTech Connect

    Not Listed

    2012-11-01

    The Advanced Fuels Campaign (AFC) under the Fuel Cycle Research and Development (FCRD) program is responsible for developing fuels technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year 2012 (FY 2012) accomplishments are highlighted below. Kemal Pasamehmetoglu is the National Technical Director for AFC.

  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. Corrosion of spent Advanced Test Reactor fuel

    SciTech Connect

    Lundberg, L.B.; Croson, M.L.

    1994-11-01

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. ARPA advanced fuel cell development

    SciTech Connect

    Dubois, L.H.

    1995-08-01

    Fuel cell technology is currently being developed at the Advanced Research Projects Agency (ARPA) for several Department of Defense applications where its inherent advantages such as environmental compatibility, high efficiency, and low noise and vibration are overwhelmingly important. These applications range from man-portable power systems of only a few watts output (e.g., for microclimate cooling and as direct battery replacements) to multimegawatt fixed base systems. The ultimate goal of the ARPA program is to develop an efficient, low-temperature fuel cell power system that operates directly on a military logistics fuel (e.g., DF-2 or JP-8). The absence of a fuel reformer will reduce the size, weight, cost, and complexity of such a unit as well as increase its reliability. In order to reach this goal, ARPA is taking a two-fold, intermediate time-frame approach to: (1) develop a viable, low-temperature proton exchange membrane (PEM) fuel cell that operates directly on a simple hydrocarbon fuel (e.g., methanol or trimethoxymethane) and (2) demonstrate a thermally integrated fuel processor/fuel cell power system operating on a military logistics fuel. This latter program involves solid oxide (SOFC), molten carbonate (MCFC), and phosphoric acid (PAFC) fuel cell technologies and concentrates on the development of efficient fuel processors, impurity scrubbers, and systems integration. A complementary program to develop high performance, light weight H{sub 2}/air PEM and SOFC fuel cell stacks is also underway. Several recent successes of these programs will be highlighted.

  15. Advanced development: Fuels

    NASA Astrophysics Data System (ADS)

    Ramohalli, K.

    1981-05-01

    The solar thermal fuels and chemicals program at Jet Propulsion Laboratory are described. High technology is developed and applied to displace fossil fuel (oil) use in the production/processing of valuable fuels and chemicals. The technical and economic feasibility is demonstrated to extent that enables the industry to participate and commercialize the product. A representative process, namely Furfural production with a bottoming of acetone, butanol and ethanol, is described. Experimental data from all solar production of furfural is discussed. Estimates are given to show the attractiveness of this process, considering its flexibility to be adaptable to dishes, troughs or central receivers. Peat, lignite and low rank coal processing, heavy oil stripping and innovative technologies for process diagnostics and control are mentioned as examples of current projects under intensive development.

  16. Advanced development: Fuels

    NASA Technical Reports Server (NTRS)

    Ramohalli, K.

    1981-01-01

    The solar thermal fuels and chemicals program at Jet Propulsion Laboratory are described. High technology is developed and applied to displace fossil fuel (oil) use in the production/processing of valuable fuels and chemicals. The technical and economic feasibility is demonstrated to extent that enables the industry to participate and commercialize the product. A representative process, namely Furfural production with a bottoming of acetone, butanol and ethanol, is described. Experimental data from all solar production of furfural is discussed. Estimates are given to show the attractiveness of this process, considering its flexibility to be adaptable to dishes, troughs or central receivers. Peat, lignite and low rank coal processing, heavy oil stripping and innovative technologies for process diagnostics and control are mentioned as examples of current projects under intensive development.

  17. 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.

  18. Advanced fuel cell development

    NASA Astrophysics Data System (ADS)

    Pierce, R. D.; Baumert, B.; Claar, T. D.; Fousek, R. J.; Huang, H. S.; Kaun, T. D.; Krumpelt, M.; Minh, N.; Mrazek, F. C.; Poeppel, R. B.

    1985-01-01

    Fuel cell research and development activities at Argonne National Laboratory (ANL) during the period January through March 1984 are described. These efforts have been directed principally toward seeking alternative cathode materials to NiO for molten carbonate fuel cells. Based on an investigation of the thermodynamically stable phases formed under cathode conditions, a number of prospective alternative cathode materials have been identified. From the list of candidates, LiFeO2, Li2MnO3, and ZnO were selected for further investigation. During this quarter, they were doped to promote conductivity and tested for solubility and ion migration in the cell environment. An investigation directed to understanding in cell densification of anode materials was initiated. In addition, calculations were made to evaluate the practicality of controlling sulfur accumulation in molten carbonate fuel cells by bleed off of a portion of the anode gas that could be recycled to the cathode. In addition, a model is being developed to predict the performance of solid oxide fuel cells as a function of cell design and operation.

  19. Advanced fuel chemistry for advanced engines.

    SciTech Connect

    Taatjes, Craig A.; Jusinski, Leonard E.; Zador, Judit; Fernandes, Ravi X.; Miller, James A.

    2009-09-01

    Autoignition chemistry is central to predictive modeling of many advanced engine designs that combine high efficiency and low inherent pollutant emissions. This chemistry, and especially its pressure dependence, is poorly known for fuels derived from heavy petroleum and for biofuels, both of which are becoming increasingly prominent in the nation's fuel stream. We have investigated the pressure dependence of key ignition reactions for a series of molecules representative of non-traditional and alternative fuels. These investigations combined experimental characterization of hydroxyl radical production in well-controlled photolytically initiated oxidation and a hybrid modeling strategy that linked detailed quantum chemistry and computational kinetics of critical reactions with rate-equation models of the global chemical system. Comprehensive mechanisms for autoignition generally ignore the pressure dependence of branching fractions in the important alkyl + O{sub 2} reaction systems; however we have demonstrated that pressure-dependent 'formally direct' pathways persist at in-cylinder pressures.

  20. 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.

  1. Advanced thermally stable jet fuels

    SciTech Connect

    Schobert, H.H.

    1999-01-31

    The Pennsylvania State University program in advanced thermally stable coal-based jet fuels has five broad objectives: (1) Development of mechanisms of degradation and solids formation; (2) Quantitative measurement of growth of sub-micrometer and micrometer-sized particles suspended in fuels during thermal stressing; (3) Characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) Elucidation of the role of additives in retarding the formation of carbonaceous solids; (5) Assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal. Future high-Mach aircraft will place severe thermal demands on jet fuels, requiring the development of novel, hybrid fuel mixtures capable of withstanding temperatures in the range of 400--500 C. In the new aircraft, jet fuel will serve as both an energy source and a heat sink for cooling the airframe, engine, and system components. The ultimate development of such advanced fuels requires a thorough understanding of the thermal decomposition behavior of jet fuels under supercritical conditions. Considering that jet fuels consist of hundreds of compounds, this task must begin with a study of the thermal degradation behavior of select model compounds under supercritical conditions. The research performed by The Pennsylvania State University was focused on five major tasks that reflect the objectives stated above: Task 1: Investigation of the Quantitative Degradation of Fuels; Task 2: Investigation of Incipient Deposition; Task 3: Characterization of Solid Gums, Sediments, and Carbonaceous Deposits; Task 4: Coal-Based Fuel Stabilization Studies; and Task 5: Exploratory Studies on the Direct Conversion of Coal to High Quality Jet Fuels. The major findings of each of these tasks are presented in this executive summary. A description of the sub-tasks performed under each of these tasks and the findings of those studies are provided in the remainder of this volume

  2. ADVANCED FUELS CAMPAIGN 2013 ACCOMPLISHMENTS

    SciTech Connect

    Not Listed

    2013-10-01

    The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This includes development of a state-of-the art Research and Development (R&D) infrastructure to support the use of “goal-oriented science-based approach.” In support of the Fuel Cycle Research and Development (FCRD) program, AFC is responsible for developing advanced fuels technologies to support the various fuel cycle options defined in the Department of Energy (DOE) Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. Accomplishments made during fiscal year (FY) 2013 are highlighted in this report, which focuses on completed work and results. The process details leading up to the results are not included; however, the technical contact is provided for each section.

  3. Advanced Nuclear Fuel Cycle Options

    SciTech Connect

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

    2010-06-01

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

  4. Advanced Fuels Campaign Execution Plan

    SciTech Connect

    Kemal Pasamehmetoglu

    2010-10-01

    The purpose of the Advanced Fuels Campaign (AFC) Execution Plan is to communicate the structure and management of research, development, and demonstration (RD&D) activities within the Fuel Cycle Research and Development (FCRD) program. Included in this document is an overview of the FCRD program, a description of the difference between revolutionary and evolutionary approaches to nuclear fuel development, the meaning of science-based development of nuclear fuels, and the “Grand Challenge” for the AFC that would, if achieved, provide a transformational technology to the nuclear industry in the form of a high performance, high reliability nuclear fuel system. The activities that will be conducted by the AFC to achieve success towards this grand challenge are described and the goals and milestones over the next 20 to 40 year period of research and development are established.

  5. Advanced Fuels Campaign Execution Plan

    SciTech Connect

    Kemal Pasamehmetoglu

    2011-09-01

    The purpose of the Advanced Fuels Campaign (AFC) Execution Plan is to communicate the structure and management of research, development, and demonstration (RD&D) activities within the Fuel Cycle Research and Development (FCRD) program. Included in this document is an overview of the FCRD program, a description of the difference between revolutionary and evolutionary approaches to nuclear fuel development, the meaning of science-based development of nuclear fuels, and the 'Grand Challenge' for the AFC that would, if achieved, provide a transformational technology to the nuclear industry in the form of a high performance, high reliability nuclear fuel system. The activities that will be conducted by the AFC to achieve success towards this grand challenge are described and the goals and milestones over the next 20 to 40 year period of research and development are established.

  6. Advanced optical document security elements

    NASA Astrophysics Data System (ADS)

    Škereš, Marek; Svoboda, Jakub; Possolt, Martin; Květoš, Milan; Fiala, Pavel

    2012-01-01

    ABSTRACT Synthetic diffractive structures represent an important tool in the optical document security. Their macroscopic visual behavior is based on properties of a very fine micro-structure which cannot be copied using common copying techniques. The visual effects can be easily observed by a common observer without any special inspection tools. However, when a high level of security is needed, additional features are often included based on an optical encryption of information. In this paper, a novel encryption technique is presented, which is based on utilizing the plastic holographic foil as a waveguide and special diffractive structures as coupling elements. When an in-coupling area is illuminated with a defined light beam, the light is coupled into the waveguide and travels to an out-coupling part. The encrypted information is encoded either in the shape of the out-coupling area or it can be formed from an out-coupling hologram in free space above the element. Both laser and normal white light sources can be used for reading the information. The coupling areas can be mixed with diffractive micro-structures forming visual effects and can be invisible during a normal observation of the hologram. The couplers can be realized using the technology fully compatible with the standard process for mastering and replication of the security elements. Several extensions of the described idea of waveguide cryptograms are also included. Finally, a set of real samples of the security elements is presented, which were realized using an advanced matrix laser lithography technique.

  7. FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Abbott, W.E.; Balent, R.

    1958-09-16

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

  8. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  9. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  10. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  11. 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.

  12. 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.

  13. Nuclear propulsion technology advanced fuels technology

    NASA Technical Reports Server (NTRS)

    Stark, Walter A., Jr.

    1993-01-01

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

  14. Advanced Thermally Stable Jet Fuels

    SciTech Connect

    A. Boehman; C. Song; H. H. Schobert; M. M. Coleman; P. G. Hatcher; S. Eser

    1998-01-01

    The Penn State program in advanced thermally stable jet fuels has five components: 1) development of mechanisms of degradation and solids formation; 2) quantitative measurement of growth of sub-micrometer and micrometer-sized particles during thermal stressing; 3) characterization of carbonaceous deposits by various instrumental and microscopic methods; 4) elucidation of the role of additives in retarding the formation of carbonaceous solids; and 5) assessment of the potential of producing high yields of cycloalkanes and hydroaromatics from coal.

  15. 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)

  16. 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.

  17. Advanced Fuels Campaign FY 2015 Accomplishments Report

    SciTech Connect

    Braase, Lori Ann; Carmack, William Jonathan

    2015-10-29

    The mission of the Advanced Fuels Campaign (AFC) is to perform research, development, and demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This report is a compilation of technical accomplishment summaries for FY-15. Emphasis is on advanced accident-tolerant LWR fuel systems, advanced transmutation fuels technologies, and capability development.

  18. Advanced spacecraft fuel cell systems

    NASA Technical Reports Server (NTRS)

    Thaller, L. H.

    1972-01-01

    The development and characteristics of advanced spacecraft fuel cell systems are discussed. The system is designed to operate on low pressure, propulsion grade hydrogen and oxygen. The specific goals are 10,000 hours of operation with refurbishment, 20 pounds per kilowatt at a sustained power of 7 KW, and 21 KW peaking capability for durations of two hours. The system rejects waste heat to the spacecraft cooling system at power levels up to 7 KW. At higher powers, the system automatically transfers to open cycle operation with overboard steam venting.

  19. 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.

  20. 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.

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. Waste Management Planned for the Advanced Fuel Cycle Facility

    SciTech Connect

    Soelberg

    2007-09-01

    The U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program has been proposed to develop and employ advanced technologies to increase the proliferation resistance of spent nuclear fuels, recover and reuse nuclear fuel resources, and reduce the amount of wastes requiring permanent geological disposal. In the initial GNEP fuel cycle concept, spent nuclear fuel is to be reprocessed to separate re-useable transuranic elements and uranium from waste fission products, for fabricating new fuel for fast reactors. The separated wastes would be converted to robust waste forms for disposal. The Advanced Fuel Cycle Facility (AFCF) is proposed by DOE for developing and demonstrating spent nuclear fuel recycling technologies and systems. The AFCF will include capabilities for receiving and reprocessing spent fuel and fabricating new nuclear fuel from the reprocessed spent fuel. Reprocessing and fuel fabrication activities will generate a variety of radioactive and mixed waste streams. Some of these waste streams are unique and unprecedented. The GNEP vision challenges traditional U.S. radioactive waste policies and regulations. Product and waste streams have been identified during conceptual design. Waste treatment technologies have been proposed based on the characteristics of the waste streams and the expected requirements for the final waste forms. Results of AFCF operations will advance new technologies that will contribute to safe and economical commercial spent fuel reprocessing facilities needed to meet the GNEP vision. As conceptual design work and research and design continues, the waste management strategies for the AFCF are expected to also evolve.

  7. 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.

  8. Chemical Kinetic Modeling of Advanced Transportation Fuels

    SciTech Connect

    PItz, W J; Westbrook, C K; Herbinet, O

    2009-01-20

    Development of detailed chemical kinetic models for advanced petroleum-based and nonpetroleum based fuels is a difficult challenge because of the hundreds to thousands of different components in these fuels and because some of these fuels contain components that have not been considered in the past. It is important to develop detailed chemical kinetic models for these fuels since the models can be put into engine simulation codes used for optimizing engine design for maximum efficiency and minimal pollutant emissions. For example, these chemistry-enabled engine codes can be used to optimize combustion chamber shape and fuel injection timing. They also allow insight into how the composition of advanced petroleum-based and non-petroleum based fuels affect engine performance characteristics. Additionally, chemical kinetic models can be used separately to interpret important in-cylinder experimental data and gain insight into advanced engine combustion processes such as HCCI and lean burn engines. The objectives are: (1) Develop detailed chemical kinetic reaction models for components of advanced petroleum-based and non-petroleum based fuels. These fuels models include components from vegetable-oil-derived biodiesel, oil-sand derived fuel, alcohol fuels and other advanced bio-based and alternative fuels. (2) Develop detailed chemical kinetic reaction models for mixtures of non-petroleum and petroleum-based components to represent real fuels and lead to efficient reduced combustion models needed for engine modeling codes. (3) Characterize the role of fuel composition on efficiency and pollutant emissions from practical automotive engines.

  9. 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.

  10. 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.

  11. Identification of leaking TRIGA fuel elements

    SciTech Connect

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

    1990-07-01

    The 100 kW TRIGA Mark I Nuclear Reactor at the University of Utah achieved initial criticality in October, 1975. Previously irradiated fuel consisting of stainless-steel- and aluminum-clad elements was acquired from the University of Arizona and the U.S. Army's Harry Diamond Laboratories in Adelphi, Maryland. Past core configurations have been comprised of both types of fuel with the aluminum-clad elements normally restricted to outer hexagonal rings of the core to provide a large safety margin between actual fuel temperature and limits set forth in the facility Technical Specifications. On October 20, 1987, trace cesium-137 contamination was discovered during routine analysis of the ion-exchange resin in the demineralizer circuit. The presence of Cs-137 indicated a possible clad defect resulting in the leakage of fission products. Reactor operations were allowed only to assist in identifying the source of the leakage. Pool water samples obtained following a two-hour operation at full power were spectroscopically analyzed and found to contain very small amounts of short-lived noble gases (e.g., Kr-85m, Kr-87, Kr-88, Xe-138) and their decay daughter products (e.g., Rb-88, Cs-138). Samples of the gaseous effluent from the facility collected in activated charcoal canisters showed no indication of fission product contamination. The small amount of activity released to the pool water suggested that a single defective element was responsible for the leakage. The instrumented fuel element and the aluminum-clad fuel were initially suspected as sources of the leakage. A simple scheme was devised to identify the defective element by exchanging four or five elements from the core with fuel in storage and then operating the reactor at 90 kW power for two hours. A pool water sample was then taken and analyzed to determine if the damaged element had been removed from the core. This process was repeated several times until all of the aluminum-clad fuel and several stainless

  12. 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.

  13. Uncertainty Analyses of Advanced Fuel Cycles

    SciTech Connect

    Laurence F. Miller; J. Preston; G. Sweder; T. Anderson; S. Janson; M. Humberstone; J. MConn; J. Clark

    2008-12-12

    The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development.

  14. 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.

  15. Advanced ceramic cladding for water reactor fuel

    SciTech Connect

    Feinroth, H.

    2000-07-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of {approximately}60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies {ge}50% would be examined.

  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. 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.

  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. Advanced Fuel Cycle Economic Sensitivity Analysis

    SciTech Connect

    David Shropshire; Kent Williams; J.D. Smith; Brent Boore

    2006-12-01

    A fuel cycle economic analysis was performed on four fuel cycles to provide a baseline for initial cost comparison using the Gen IV Economic Modeling Work Group G4 ECON spreadsheet model, Decision Programming Language software, the 2006 Advanced Fuel Cycle Cost Basis report, industry cost data, international papers, the nuclear power related cost study from MIT, Harvard, and the University of Chicago. The analysis developed and compared the fuel cycle cost component of the total cost of energy for a wide range of fuel cycles including: once through, thermal with fast recycle, continuous fast recycle, and thermal recycle.

  3. Advanced Fuels Campaign FY 2011 Accomplishments Report

    SciTech Connect

    Not Listed

    2011-11-01

    One of the major research and development (R&D) areas under the Fuel Cycle Research and Development (FCRD) program is advanced fuels development. The Advanced Fuels Campaign (AFC) has the responsibility to develop advanced fuel technologies for the Department of Energy (DOE) using a science-based approach focusing on developing a microstructural understanding of nuclear fuels and materials. Accomplishments made during fiscal year (FY 20) 2011 are highlighted in this report, which focuses on completed work and results. The process details leading up to the results are not included; however, the technical contact is provided for each section. The order of the accomplishments in this report is consistent with the AFC work breakdown structure (WBS).

  4. Criticality Safety Evaluation for the Advanced Test Reactor U-Mo Demonstration Elements

    SciTech Connect

    Leland M. Montierth

    2010-12-01

    The Reduced Enrichment Research Test Reactors (RERTR) fuel development program is developing a high uranium density fuel based on a (LEU) uranium-molybdenum alloy. Testing of prototypic RERTR fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. Two RERTR-Full Size Demonstration fuel elements based on the ATR-Reduced YA elements (all but one plate fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). The two fuel elements will be irradiated in alternating cycles such that only one element is loaded in the reactor at a time. Existing criticality analyses have analyzed Standard (HEU) ATR elements (all plates fueled) from which controls have been derived. This criticality safety evaluation (CSE) documents analysis that determines the reactivity of the Demonstration fuel elements relative to HEU ATR elements and shows that the Demonstration elements are bound by the Standard HEU ATR elements and existing HEU ATR element controls are applicable to the Demonstration elements.

  5. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels. PMID:23274823

  6. Advanced Biorefineries for Production of Fuel Ethanol

    Technology Transfer Automated Retrieval System (TEKTRAN)

    This review, "Advanced biorefineries for production of fuel ethanol," is a chapter in the Wiley book entitled Biomass to Biofuels: Strategies for Global Industries and is intended to cover all major ethanol production processes to date. The chapter discusses current fuel ethanol production processe...

  7. Physics challenges for advanced fuel cycle assessment

    SciTech Connect

    Giuseppe Palmiotti; Massimo Salvatores; Gerardo Aliberti

    2014-06-01

    Advanced fuel cycles and associated optimized reactor designs will require substantial improvements in key research area to meet new and more challenging requirements. The present paper reviews challenges and issues in the field of reactor and fuel cycle physics. Typical examples are discussed with, in some cases, original results.

  8. 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.

  9. 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.

  10. 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.

  11. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    SciTech Connect

    Not Listed

    2013-03-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) organized a Cladding and Coatings operational meeting February 12-13, 2013, at Oak Ridge National Laboratory (ORNL). Representatives from the U.S. Department of Energy (DOE), national laboratories, industry, and universities attended the two-day meeting. The purpose of the meeting was to discuss advanced cladding and cladding coating research and development (R&D); review experimental testing capabilities for assessing accident tolerant fuels; and review industry/university plans and experience in light water reactor (LWR) cladding and coating R&D.

  12. Advanced Fuels Campaign FY 2010 Accomplishments Report

    SciTech Connect

    Lori Braase

    2010-12-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) Accomplishment Report documents the high-level research and development results achieved in fiscal year 2010. The AFC program has been given responsibility to develop advanced fuel technologies for the Department of Energy (DOE) using a science-based approach focusing on developing a microstructural understanding of nuclear fuels and materials. The science-based approach combines theory, experiments, and multi-scale modeling and simulation aimed at a fundamental understanding of the fuel fabrication processes and fuel and clad performance under irradiation. The scope of the AFC includes evaluation and development of multiple fuel forms to support the three fuel cycle options described in the Sustainable Fuel Cycle Implementation Plan4: Once-Through Cycle, Modified-Open Cycle, and Continuous Recycle. The word “fuel” is used generically to include fuels, targets, and their associated cladding materials. This document includes a brief overview of the management and integration activities; but is primarily focused on the technical accomplishments for FY-10. Each technical section provides a high level overview of the activity, results, technical points of contact, and applicable references.

  13. Advanced-fuel-cell development

    NASA Astrophysics Data System (ADS)

    Pierce, R. D.; Arons, R. M.; Dusek, J. T.; Fraioli, A. V.; Kucera, G. H.; Sim, J. W.; Smith, J. L.

    1982-08-01

    Fuel cell research and development activities are described. The efforts are directed toward: (1) understanding of component behavior in molten carbonate fuel cells, and (2) developing alternative concepts for components. The principal focus was on the development of sintered gamma LiAlO2 electrolyte supports, stable NiO cathodes, and hydrogen diffusion barriers. Cell tests were performed to assess diffusion barriers and to study cathode voltage relaxation following current interruption.

  14. 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.

  15. 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.

  16. 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)

  17. Advanced-fuel-cell development

    NASA Astrophysics Data System (ADS)

    Pierce, R. D.; Arons, R. M.; Dusek, J. T.; Fraioli, A. V.; Kucera, G. H.; Sim, J. W.; Smith, J. L.

    1982-06-01

    The fuel cell research and development activities at Argonne National Laboratory (ANL) for the period October through December 1980. These efforts have been directed toward (1) developing alternative concepts for components of molten carbonate fuel cells, and (2) improving understanding of component behavior. The principal focus has been on development of gamma-LiAlO2 sinters as electrolyte structures. Green bodies were prepared by tape casting and then sintering beta-LiAlO2; this has produced gamma-LiAlO2 sinters of 69% porosity. In addition, a cathode prepared by sintering lithiated nickel oxide was tested in a 10-cm square cell.

  18. 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)

  19. Elements of an advanced integrated operator control station

    SciTech Connect

    Clarke, M.M.; Kreifeldt, J.G.

    1984-01-01

    One of the critical determinants of peformance for any remotely operated maintenance system is the compatibility achieved between elements of the man/machine interface (e.g., master manipulator controller, controls, displays, etc.) and the human operator. In the Remote Control Engineering task of the Consolidated Fuel Reprocessing Program, considerable attention has been devoted to optimizing the man/machine interface of the operator control station. This system must be considered an integral element of the overall maintenance work system which includes transporters, manipulators, remote viewing, and other parts. The control station must reflect the integration of the operator team, control/display panels, manipulator master controllers, and remote viewing monitors. Human factors principles and experimentation have been used in the development of an advanced integrated operator control station designed for the advance servomanipulator. Key features of this next-generation design are summarized in this presentation. 7 references, 4 figures.

  20. Elements of an advanced integrated operator control station

    SciTech Connect

    Clarke, M.M.; Kreifeldt, J.G.

    1984-01-01

    One of the critical determinants of performance for any remotely operated maintenance system is the compatibility achieved between elements of the man/machine interface (e.g., master manipulator controller, controls, displays) and the human operator. In the remote control engineering task of the Consolidated Fuel Reprocessing Program, considerable attention has been devoted to optimizing the man/machine interface of the operator control station. This system must be considered an integral element of the overall maintenance work system which includes transporters, manipulators, remote viewing, and other parts. The control station must reflect the integration of the operator team, control/display panels, manipulator master controllers, and remote viewing monitors. Human factors principles and experimentation have been used in the development of an advanced integrated operator control station designed for the advance servomanipulator. Key features of this next-generation design are summarized in this presentation. 7 references, 4 figures.

  1. Thermochemical modelling of advanced CANDU reactor fuel

    NASA Astrophysics Data System (ADS)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  2. Microscale microbial fuel cells: Advances and challenges.

    PubMed

    Choi, Seokheun

    2015-07-15

    The next generation of sustainable energy could come from microorganisms; evidence that it can be seen with the given rise of Electromicrobiology, the study of microorganisms' electrical properties. Many recent advances in electromicrobiology stem from studying microbial fuel cells (MFCs), which are gaining acceptance as a future alternative "green" energy technology and energy-efficient wastewater treatment method. MFCs are powered by living microorganisms with clean and sustainable features; they efficiently catalyse the degradation of a broad range of organic substrates under natural conditions. There is also increasing interest in photosynthetic MFCs designed to harness Earth's most abundant and promising energy source (solar irradiation). Despite their vast potential and promise, however, MFCs and photosynthetic MFCs have not yet successfully translated into commercial applications because they demonstrate persistent performance limitations and bottlenecks associated with scaling up. Instead, microscale MFCs have received increasing attention as a unique platform for various applications such as powering small portable electronic elements in remote locations, performing fundamental studies of microorganisms, screening bacterial strains, and toxicity detection in water. Furthermore, the stacking of miniaturized MFCs has been demonstrated to offer larger power densities than a single macroscale MFC in terms of scaling up. In this overview, we discuss recent achievements in microscale MFCs as well as their potential applications. Further scientific and technological challenges are also reviewed. PMID:25703724

  3. Future Transient Testing of Advanced Fuels

    SciTech Connect

    Jon Carmack

    2009-09-01

    The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by the

  4. Advanced diesel electronic fuel injection and turbocharging

    NASA Astrophysics Data System (ADS)

    Beck, N. J.; Barkhimer, R. L.; Steinmeyer, D. C.; Kelly, J. E.

    1993-12-01

    The program investigated advanced diesel air charging and fuel injection systems to improve specific power, fuel economy, noise, exhaust emissions, and cold startability. The techniques explored included variable fuel injection rate shaping, variable injection timing, full-authority electronic engine control, turbo-compound cooling, regenerative air circulation as a cold start aid, and variable geometry turbocharging. A Servojet electronic fuel injection system was designed and manufactured for the Cummins VTA-903 engine. A special Servojet twin turbocharger exhaust system was also installed. A series of high speed combustion flame photos was taken using the single cylinder optical engine at Michigan Technological University. Various fuel injection rate shapes and nozzle configurations were evaluated. Single-cylinder bench tests were performed to evaluate regenerative inlet air heating techniques as an aid to cold starting. An exhaust-driven axial cooling air fan was manufactured and tested on the VTA-903 engine.

  5. Advanced Catalysts for Fuel Cells

    NASA Technical Reports Server (NTRS)

    Narayanan, Sekharipuram R.; Whitacre, Jay; Valdez, T. I.

    2006-01-01

    This viewgraph presentation reviews the development of catalyst for Fuel Cells. The objectives of the project are to reduce the cost of stack components and reduce the amount of precious metal used in fuel cell construction. A rapid combinatorial screening technique based on multi-electrode thin film array has been developed and validated for identifying catalysts for oxygen reduction; focus shifted from methanol oxidation in FY05 to oxygen reduction in FY06. Multi-electrode arrays of thin film catalysts of Pt-Ni and Pt-Ni-Zr have been deposited. Pt-Ni and have been characterized electrochemically and structurally. Pt-Ni-Zr and Pt-Ni films show higher current density and onset potential compared to Pt. Electrocatalytic activity and onset potential are found to be strong function of the lattice constant. Thin film Pt(59)Ni(39)Zr(2) can provide 10 times the current density of thin film Pt. Thin film Pt(59)Ni(39)Zr(2) also shows 65mV higher onset potential than Pt.

  6. 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)

  7. 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.

  8. 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.

  9. Advanced coal-fueled gas turbine systems

    SciTech Connect

    Wenglarz, R.A.

    1994-08-01

    Several technology advances since the early coal-fueled turbine programs that address technical issues of coal as a turbine fuel have been developed in the early 1980s: Coal-water suspensions as fuel form, improved methods for removing ash and contaminants from coal, staged combustion for reducing NO{sub x} emissions from fuel-bound nitrogen, and greater understanding of deposition/erosion/corrosion and their control. Several Advanced Coal-Fueled Gas Turbine Systems programs were awarded to gas turbine manufacturers for for components development and proof of concept tests; one of these was Allison. Tests were conducted in a subscale coal combustion facility and a full-scale facility operating a coal combustor sized to the Allison Model 501-K industrial turbine. A rich-quench-lean (RQL), low nitrogen oxide combustor design incorporating hot gas cleanup was developed for coal fuels; this should also be applicable to biomass, etc. The combustor tests showed NO{sub x} and CO emissions {le} levels for turbines operating with natural gas. Water washing of vanes from the turbine removed the deposits. Systems and economic evaluations identified two possible applications for RQL turbines: Cogeneration plants based on Allison 501-K turbine (output 3.7 MW(e), 23,000 lbs/hr steam) and combined cycle power plants based on 50 MW or larger gas turbines. Coal-fueled cogeneration plant configurations were defined and evaluated for site specific factors. A coal-fueled turbine combined cycle plant design was identified which is simple, compact, and results in lower capital cost, with comparable efficiency and low emissions relative to other coal technologies (gasification, advanced PFBC).

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

    SciTech Connect

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

    2014-01-09

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

  11. Advances in fuel cell vehicle design

    NASA Astrophysics Data System (ADS)

    Bauman, Jennifer

    Factors such as global warming, dwindling fossil fuel reserves, and energy security concerns combine to indicate that a replacement for the internal combustion engine (ICE) vehicle is needed. Fuel cell vehicles have the potential to address the problems surrounding the ICE vehicle without imposing any significant restrictions on vehicle performance, driving range, or refuelling time. Though there are currently some obstacles to overcome before attaining the widespread commercialization of fuel cell vehicles, such as improvements in fuel cell and battery durability, development of a hydrogen infrastructure, and reduction of high costs, the fundamental concept of the fuel cell vehicle is strong: it is efficient, emits zero harmful emissions, and the hydrogen fuel can be produced from various renewable sources. Therefore, research on fuel cell vehicle design is imperative in order to improve vehicle performance and durability, increase efficiency, and reduce costs. This thesis makes a number of key contributions to the advancement of fuel cell vehicle design within two main research areas: powertrain design and DC/DC converters. With regards to powertrain design, this research first analyzes various powertrain topologies and energy storage system types. Then, a novel fuel cell-battery-ultracapacitor topology is presented which shows reduced mass and cost, and increased efficiency, over other promising topologies found in the literature. A detailed vehicle simulator is created in MATLAB/Simulink in order to simulate and compare the novel topology with other fuel cell vehicle powertrain options. A parametric study is performed to optimize each powertrain and general conclusions for optimal topologies, as well as component types and sizes, for fuel cell vehicles are presented. Next, an analytical method to optimize the novel battery-ultracapacitor energy storage system based on maximizing efficiency, and minimizing cost and mass, is developed. This method can be applied

  12. 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.

  13. 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.

  14. Modeling of advanced fossil fuel power plants

    NASA Astrophysics Data System (ADS)

    Zabihian, Farshid

    The first part of this thesis deals with greenhouse gas (GHG) emissions from fossil fuel-fired power stations. The GHG emission estimation from fossil fuel power generation industry signifies that emissions from this industry can be significantly reduced by fuel switching and adaption of advanced power generation technologies. In the second part of the thesis, steady-state models of some of the advanced fossil fuel power generation technologies are presented. The impacts of various parameters on the solid oxide fuel cell (SOFC) overpotentials and outputs are investigated. The detail analyses of operation of the hybrid SOFC-gas turbine (GT) cycle when fuelled with methane and syngas demonstrate that the efficiencies of the cycles with and without anode exhaust recirculation are close, but the specific power of the former is much higher. The parametric analysis of the performance of the hybrid SOFC-GT cycle indicates that increasing the system operating pressure and SOFC operating temperature and fuel utilization factor improves cycle efficiency, but the effects of the increasing SOFC current density and turbine inlet temperature are not favourable. The analysis of the operation of the system when fuelled with a wide range of fuel types demonstrates that the hybrid SOFC-GT cycle efficiency can be between 59% and 75%, depending on the inlet fuel type. Then, the system performance is investigated when methane as a reference fuel is replaced with various species that can be found in the fuel, i.e., H2, CO2, CO, and N 2. The results point out that influence of various species can be significant and different for each case. The experimental and numerical analyses of a biodiesel fuelled micro gas turbine indicate that fuel switching from petrodiesel to biodiesel can influence operational parameters of the system. The modeling results of gas turbine-based power plants signify that relatively simple models can predict plant performance with acceptable accuracy. The unique

  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. Computational Design of Advanced Nuclear Fuels

    SciTech Connect

    Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

    2014-06-03

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

  17. Development of advanced fuel cell system

    NASA Technical Reports Server (NTRS)

    Grevstad, P. E.

    1972-01-01

    Weight, life and performance characteristics optimization of hydrogen-oxygen fuel cell power systems were considered. A promising gold alloy cathode catalyst was identified and tested in a cell for 5,000 hours. The compatibility characteristics of candidate polymer structural materials were measured after exposure to electrolyte and water vapor for 8,000 hours. Lightweight cell designs were prepared and fabrication techniques to produce them were developed. Testing demonstrated that predicted performance was achieved. Lightweight components for passive product water removal and evaporative cooling of cells were demonstrated. Systems studies identified fuel cell powerplant concepts for meeting the requirements of advanced spacecraft.

  18. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    SciTech Connect

    Jon Carmack; Kemal O. Pasamehmetoglu; David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  19. 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)

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. Reactor physics analyses of the advanced neutron source three-element core

    SciTech Connect

    Gehin, J.C.

    1995-08-01

    A reactor physics analysis was performed for the Advanced Neutron Source reactor with a three-element core configuration. The analysis was performed with a two-dimensional r-z 20-energy-group finite-difference diffusion theory model of the 17-d fuel cycle. The model included equivalent r-z geometry representations of the central control rods, the irradiation and production targets, and reflector components. Calculated quantities include fuel cycle parameters, fuel element power distributions, unperturbed neutron fluxes in the reflector and target regions, reactivity perturbations, and neutron kinetics parameters.

  5. 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.

  6. 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.

  7. Fuel qualification plan for the Advanced Neutron Source Reactor

    SciTech Connect

    Copeland, G.L.

    1995-07-01

    This report describes the development and qualification plan for the fuel for the Advanced Neutron Source. The reference fuel is U{sub 3}Si{sub 2}, dispersed in aluminum and clad in 6061 aluminum. This report was prepared in May 1994, at which time the reference design was for a two-element core containing highly enriched uranium (93% {sup 235}U) . The reactor was in the process of being redesigned to accommodate lowered uranium enrichment and became a three-element core containing a higher volume fraction of uranium enriched to 50% {sup 235}U. Consequently, this report was not issued at that time and would have been revised to reflect the possibly different requirements of the lower-enrichment, higher-volume fraction fuel. Because the reactor is now being canceled, this unrevised report is being issued for archival purposes. The report describes the fabrication and inspection development plan, the irradiation tests and performance modeling to qualify performance, the transient testing that is part of the safety program, and the interactions and interfaces of the fuel development with other tasks.

  8. 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.

  9. Study of advanced fuel system concepts for commercial aircraft and engines

    NASA Technical Reports Server (NTRS)

    Versaw, E. F.; Brewer, G. D.; Byers, W. D.; Fogg, H. W.; Hanks, D. E.; Chirivella, J.

    1983-01-01

    The impact on a commercial transport aircraft of using fuels which have relaxed property limits relative to current commercial jet fuel was assessed. The methodology of the study is outlined, fuel properties are discussed, and the effect of the relaxation of fuel properties analyzed. Advanced fuel system component designs that permit the satisfactory use of fuel with the candidate relaxed properties in the subject aircraft are described. The two fuel properties considered in detail are freezing point and thermal stability. Three candidate fuel system concepts were selected and evaluated in terms of performance, cost, weight, safety, and maintainability. A fuel system that incorporates insulation and electrical heating elements on fuel tank lower surfaces was found to be most cost effective for the long term.

  10. 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.

  11. Development of advanced fuel cell system

    NASA Technical Reports Server (NTRS)

    Gitlow, B.; Meyer, A. P.; Bell, W. F.; Martin, R. E.

    1978-01-01

    An experimental program was conducted continuing the development effort to improve the weight, life, and performance characteristics of hydrogen-oxygen alkaline fuel cells for advanced power systems. These advanced technology cells operate with passive water removal which contributes to a lower system weight and extended operating life. Endurance evaluation of two single cells and two, two-cell plaques was continued. Three new test articles were fabricated and tested. A single cell completed 7038 hours of endurance testing. This cell incorporated a Fybex matrix, hybrid-frame, PPF anode, and a 90 Au/10 Pt cathode. This configuration was developed to extend cell life. Two cell plaques with dedicated flow fields and manifolds for all fluids did not exhibit the cell-to-cell electrolyte transfer that limited the operating life of earlier multicell plaques.

  12. 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.

  13. 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)

  14. Advanced Coal-Fueled Gas Turbine Program

    SciTech Connect

    Horner, M.W.; Ekstedt, E.E.; Gal, E.; Jackson, M.R.; Kimura, S.G.; Lavigne, R.G.; Lucas, C.; Rairden, J.R.; Sabla, P.E.; Savelli, J.F.; Slaughter, D.M.; Spiro, C.L.; Staub, F.W.

    1989-02-01

    The objective of the original Request for Proposal was to establish the technological bases necessary for the subsequent commercial development and deployment of advanced coal-fueled gas turbine power systems by the private sector. The offeror was to identify the specific application or applications, toward which his development efforts would be directed; define and substantiate the technical, economic, and environmental criteria for the selected application; and conduct such component design, development, integration, and tests as deemed necessary to fulfill this objective. Specifically, the offeror was to choose a system through which ingenious methods of grouping subcomponents into integrated systems accomplishes the following: (1) Preserve the inherent power density and performance advantages of gas turbine systems. (2) System must be capable of meeting or exceeding existing and expected environmental regulations for the proposed application. (3) System must offer a considerable improvement over coal-fueled systems which are commercial, have been demonstrated, or are being demonstrated. (4) System proposed must be an integrated gas turbine concept, i.e., all fuel conditioning, all expansion gas conditioning, or post-expansion gas cleaning, must be integrated into the gas turbine system.

  15. Alternative Fuel and Advanced Vehicle Tools (AFAVT), AFDC (Fact Sheet)

    SciTech Connect

    Not Available

    2010-01-01

    The Alternative Fuels and Advanced Vehicles Web site offers a collection of calculators, interactive maps, and informational tools to assist fleets, fuel providers, and others looking to reduce petroleum consumption in the transportation sector.

  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. 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.

  20. Advanced fuel system technology for utilizing broadened property aircraft fuels

    NASA Technical Reports Server (NTRS)

    Reck, G. M.

    1980-01-01

    Factors which will determine the future supply and cost of aviation turbine fuels are discussed. The most significant fuel properties of volatility, fluidity, composition, and thermal stability are discussed along with the boiling ranges of gasoline, naphtha jet fuels, kerosene, and diesel oil. Tests were made to simulate the low temperature of an aircraft fuel tank to determine fuel tank temperatures for a 9100-km flight with and without fuel heating; the effect of N content in oil-shale derived fuels on the Jet Fuel Thermal Oxidation Tester breakpoint temperature was measured. Finally, compatibility of non-metallic gaskets, sealants, and coatings with increased aromatic content jet fuels was examined.

  1. 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.

  2. Advanced membrane electrode assemblies for fuel cells

    DOEpatents

    Kim, Yu Seung; Pivovar, Bryan S.

    2012-07-24

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  3. Advanced membrane electrode assemblies for fuel cells

    SciTech Connect

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  4. 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.

  5. Advanced fuel system technology for utilizing broadened property aircraft fuels

    NASA Technical Reports Server (NTRS)

    Reck, G. M.

    1980-01-01

    Possible changes in fuel properties are identified based on current trends and projections. The effect of those changes with respect to the aircraft fuel system are examined and some technological approaches to utilizing those fuels are described.

  6. 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.

  7. Advanced coal-fueled gas turbine systems

    SciTech Connect

    Not Available

    1992-09-01

    Westinghouse's Advanced Coal-Fueled Gas Turbine System Program (DE-AC2l-86MC23167) was originally split into two major phases - a Basic Program and an Option. The Basic Program also contained two phases. The development of a 6 atm, 7 lb/s, 12 MMBtu/hr slagging combustor with an extended period of testing of the subscale combustor, was the first part of the Basic Program. In the second phase of the Basic Program, the combustor was to be operated over a 3-month period with a stationary cascade to study the effect of deposition, erosion and corrosion on combustion turbine components. The testing of the concept, in subscale, has demonstrated its ability to handle high- and low-sulfur bituminous coals, and low-sulfur subbituminous coal. Feeding the fuel in the form of PC has proven to be superior to CWM type feed. The program objectives relative to combustion efficiency, combustor exit temperature, NO[sub x] emissions, carbon burnout, and slag rejection have been met. Objectives for alkali, particulate, and SO[sub x] levels leaving the combustor were not met by the conclusion of testing at Textron. It is planned to continue this testing, to achieve all desired emission levels, as part of the W/NSP program to commercialize the slagging combustor technology.

  8. 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.

  9. Assessment of SFR fuel pin performance codes under advanced fuel for minor actinide transmutation

    SciTech Connect

    Bouineau, V.; Lainet, M.; Chauvin, N.; Pelletier, M.

    2013-07-01

    Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is, therefore, an option for the reduction of radiotoxicity and residual power packages as well as the repository area. In the SUPERFACT Experiment four different oxide fuels containing high and low concentrations of {sup 237}Np and {sup 241}Am, representing the homogeneous and heterogeneous in-pile recycling concepts, were irradiated in the PHENIX reactor. The behavior of advanced fuel materials with minor actinide needs to be fully characterized, understood and modeled in order to optimize the design of this kind of fuel elements and to evaluate its performances. This paper assesses the current predictability of fuel performance codes TRANSURANUS and GERMINAL V2 on the basis of post irradiation examinations of the SUPERFACT experiment for pins with low minor actinide content. Their predictions have been compared to measured data in terms of geometrical changes of fuel and cladding, fission gases behavior and actinide and fission product distributions. The results are in good agreement with the experimental results, although improvements are also pointed out for further studies, especially if larger content of minor actinide will be taken into account in the codes. (authors)

  10. 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

  11. 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.

  12. 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)

  13. Advanced Combustion and Fuels; NREL (National Renewable Energy Laboratory)

    SciTech Connect

    Zigler, Brad

    2015-06-08

    Presented at the U.S. Department of Energy Vehicle Technologies Office 2015 Annual Merit Review and Peer Evaluation Meeting, held June 8-12, 2015, in Arlington, Virginia. It addresses technical barriers of inadequate data and predictive tools for fuel and lubricant effects on advanced combustion engines, with the strategy being through collaboration, develop techniques, tools, and data to quantify critical fuel physico-chemical effects to enable development of advanced combustion engines that use alternative fuels.

  14. On-Going Comparison of Advanced Fuel Cycle Options

    SciTech Connect

    Piet, S.J.; Bennett, R.G.; Dixon, B.W.; Herring, J.S.; Shropshire, D.E.; Roth, M.; Smith, J.D.; Finck, P.; Hill, R.; Laidler, J.; Pasamehmetoglu, K.

    2004-10-03

    This paper summarizes the current comprehensive comparison of four major fuel cycle strategies: once-through, thermal recycle, thermal+fast recycle, fast recycle. It then proceeds to summarize comparison of the major technology options for the key elements of the fuel cycle that can implement each of the four strategies - separation processing, transmutation reactors, and fuels.

  15. 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.

  16. 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.

  17. 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.

  18. Elemental advances of ultraheavy cosmic rays

    NASA Technical Reports Server (NTRS)

    1984-01-01

    The elemental composition of the cosmic-ray source is different from that which has been generally taken as the composition of the solar system. No general enrichment of products of either r-process or s-process nucleosynthesis accounts for the differences over the entire range of ultraheavy (Z 30) elements; specific determination of nucleosynthetic contributions to the differences depends upon an understanding of the nature of any acceleration fractionation. Comparison between the cosmic-ray source abundances and the abundances of C1 and C2 chondritic meteorites suggests that differences between the cosmic-ray source and the standard (C1) solar system may not be due to acceleration fractionation of the cosmic rays, but rather to a fractionation of the C1 abundances with respect to the interstellar abundances.

  19. 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.

  20. 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.

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

    SciTech Connect

    Jon Carmack

    2014-01-01

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

  2. 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.

  3. 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)

  4. Metallic fuels: The EBR-II legacy and recent advances

    SciTech Connect

    Douglas L. Porter; Steven L. Hayes; J. Rory Kennedy

    2012-09-01

    Experimental Breeder Reactor – II (EBR-II) metallic fuel was qualified for high burnup to approximately 10 atomic per cent. Subsequently, the electrometallurgical treatment of this fuel was demonstrated. Advanced metallic fuels are now investigated for increased performance, including ultra-high burnup and actinide burning. Advances include additives to mitigate the fuel/cladding chemical interaction and uranium alloys that combine Mo, Ti and Zr to improve alloy performance. The impacts of the advances—on fabrication, waste streams, electrorefining, etc.—are found to be minimal and beneficial. Owing to extensive research literature and computational methods, only a modest effort is required to complete their development.

  5. Advanced Fuel Cell System Thermal Management for NASA Exploration Missions

    NASA Technical Reports Server (NTRS)

    Burke, Kenneth A.

    2009-01-01

    The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA exploration program. An analysis of a state-of-the-art fuel cell cooling systems was done to benchmark the portion of a fuel cell system s mass that is dedicated to thermal management. Additional analysis was done to determine the key performance targets of the advanced passive thermal management technology that would substantially reduce fuel cell system mass.

  6. 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.

  7. The DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program

    SciTech Connect

    David Petti; Hans Gougar; Gary Bell

    2005-05-01

    The Department of Energy has established the Advanced Gas Reactor Fuel Development and Qualification Program to address the following overall goals: Provide a baseline fuel qualification data set in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Gas-reactor fuel performance demonstration and qualification comprise the longest duration research and development (R&D) task for the NGNP feasibility. The baseline fuel form is to be demonstrated and qualified for a peak fuel centerline temperature of 1250°C. Support near-term deployment of an NGNP by reducing market entry risks posed by technical uncertainties associated with fuel production and qualification. Utilize international collaboration mechanisms to extend the value of DOE resources. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, postirradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete fundamental understanding of the relationship between the fuel fabrication process, key fuel properties, the irradiation performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. Fuel performance modeling and analysis of the fission product behavior in the primary circuit are important aspects of this work. The performance models are considered essential for several reasons, including guidance for the plant designer in establishing the core design and operating limits, and demonstration to the licensing authority that the applicant has a thorough understanding of the in-service behavior of the fuel system. The fission product behavior task will also provide primary source term data needed for licensing. An overview of the program and recent progress will be presented.

  8. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    SciTech Connect

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  9. 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.

  10. 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.

  11. 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.

  12. 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}.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Advances in 3D electromagnetic finite element modeling

    SciTech Connect

    Nelson, E.M.

    1997-08-01

    Numerous advances in electromagnetic finite element analysis (FEA) have been made in recent years. The maturity of frequency domain and eigenmode calculations, and the growth of time domain applications is briefly reviewed. A high accuracy 3D electromagnetic finite element field solver employing quadratic hexahedral elements and quadratic mixed-order one-form basis functions will also be described. The solver is based on an object-oriented C++ class library. Test cases demonstrate that frequency errors less than 10 ppm can be achieved using modest workstations, and that the solutions have no contamination from spurious modes. The role of differential geometry and geometrical physics in finite element analysis is also discussed.

  18. 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)

  19. Advanced Fuels Campaign FY 2014 Accomplishments Report

    SciTech Connect

    Lori Braase; W. Edgar May

    2014-10-01

    The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance.

  20. Advanced Safeguards Approaches for New TRU Fuel Fabrication Facilities

    SciTech Connect

    Durst, Philip C.; Ehinger, Michael H.; Boyer, Brian; Therios, Ike; Bean, Robert; Dougan, A.; Tolk, K.

    2007-12-15

    This second report in a series of three reviews possible safeguards approaches for the new transuranic (TRU) fuel fabrication processes to be deployed at AFCF – specifically, the ceramic TRU (MOX) fuel fabrication line and the metallic (pyroprocessing) line. The most common TRU fuel has been fuel composed of mixed plutonium and uranium dioxide, referred to as “MOX”. However, under the Advanced Fuel Cycle projects custom-made fuels with higher contents of neptunium, americium, and curium may also be produced to evaluate if these “minor actinides” can be effectively burned and transmuted through irradiation in the ABR. A third and final report in this series will evaluate and review the advanced safeguards approach options for the ABR. In reviewing and developing the advanced safeguards approach for the new TRU fuel fabrication processes envisioned for AFCF, the existing international (IAEA) safeguards approach at the Plutonium Fuel Production Facility (PFPF) and the conceptual approach planned for the new J-MOX facility in Japan have been considered as a starting point of reference. The pyro-metallurgical reprocessing and fuel fabrication process at EBR-II near Idaho Falls also provided insight for safeguarding the additional metallic pyroprocessing fuel fabrication line planned for AFCF.

  1. Advanced fuel cell concepts for future NASA missions

    NASA Astrophysics Data System (ADS)

    Stedman, J. K.

    1987-09-01

    Studies of primary fuel cells for advanced all electric shuttle type vehicles show an all fuel cell power system with peak power capability of 100's of kW to be potentially lighter and have lower life cycle costs than a hybrid system using advanced H2O2 APU's for peak power and fuel cells for low power on orbit. Fuel cell specific weights of 1 to 3 lb/kW, a factor of 10 improvement over the orbiter power plant, are projected for the early 1990's. For satellite applications, a study to identify high performance regenerative hydrogen oxygen fuel cell concepts for geosynchronous orbit was completed. Emphasis was placed on concepts with the potential for high energy density (Wh/lb) and passive means for water and heat management to maximize system reliability. Both alkaline electrolyte and polymer membrane fuel cells were considered.

  2. Advanced fuel cell concepts for future NASA missions

    NASA Technical Reports Server (NTRS)

    Stedman, J. K.

    1987-01-01

    Studies of primary fuel cells for advanced all electric shuttle type vehicles show an all fuel cell power system with peak power capability of 100's of kW to be potentially lighter and have lower life cycle costs than a hybrid system using advanced H2O2 APU's for peak power and fuel cells for low power on orbit. Fuel cell specific weights of 1 to 3 lb/kW, a factor of 10 improvement over the orbiter power plant, are projected for the early 1990's. For satellite applications, a study to identify high performance regenerative hydrogen oxygen fuel cell concepts for geosynchronous orbit was completed. Emphasis was placed on concepts with the potential for high energy density (Wh/lb) and passive means for water and heat management to maximize system reliability. Both alkaline electrolyte and polymer membrane fuel cells were considered.

  3. Study of advanced fuel system concepts for commercial aircraft

    NASA Technical Reports Server (NTRS)

    Coffinberry, G. A.

    1985-01-01

    An analytical study was performed in order to assess relative performance and economic factors involved with alternative advanced fuel systems for future commercial aircraft operating with broadened property fuels. The DC-10-30 wide-body tri-jet aircraft and the CF6-8OX engine were used as a baseline design for the study. Three advanced systems were considered and were specifically aimed at addressing freezing point, thermal stability and lubricity fuel properties. Actual DC-10-30 routes and flight profiles were simulated by computer modeling and resulted in prediction of aircraft and engine fuel system temperatures during a nominal flight and during statistical one-day-per-year cold and hot flights. Emergency conditions were also evaluated. Fuel consumption and weight and power extraction results were obtained. An economic analysis was performed for new aircraft and systems. Advanced system means for fuel tank heating included fuel recirculation loops using engine lube heat and generator heat. Environmental control system bleed air heat was used for tank heating in a water recirculation loop. The results showed that fundamentally all of the three advanced systems are feasible but vary in their degree of compatibility with broadened-property fuel.

  4. 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)

  5. 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)

  6. Advance finite element modeling of rotor blade aeroelasticity

    NASA Technical Reports Server (NTRS)

    Straub, F. K.; Sangha, K. B.; Panda, B.

    1994-01-01

    An advanced beam finite element has been developed for modeling rotor blade dynamics and aeroelasticity. This element is part of the Element Library of the Second Generation Comprehensive Helicopter Analysis System (2GCHAS). The element allows modeling of arbitrary rotor systems, including bearingless rotors. It accounts for moderately large elastic deflections, anisotropic properties, large frame motion for maneuver simulation, and allows for variable order shape functions. The effects of gravity, mechanically applied and aerodynamic loads are included. All kinematic quantities required to compute airloads are provided. In this paper, the fundamental assumptions and derivation of the element matrices are presented. Numerical results are shown to verify the formulation and illustrate several features of the element.

  7. 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.

  8. 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.

  9. 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.

  10. Advances in direct oxidation methanol fuel cells

    NASA Technical Reports Server (NTRS)

    Surampudi, S.; Narayanan, S. R.; Vamos, E.; Frank, H.; Halpert, G.; Laconti, Anthony B.; Kosek, J.; Prakash, G. K. Surya; Olah, G. A.

    1993-01-01

    Fuel cells that can operate directly on fuels such as methanol are attractive for low to medium power applications in view of their low weight and volume relative to other power sources. A liquid feed direct methanol fuel cell has been developed based on a proton exchange membrane electrolyte and Pt/Ru and Pt catalyzed fuel and air/O2 electrodes, respectively. The cell has been shown to deliver significant power outputs at temperatures of 60 to 90 C. The cell voltage is near 0.5 V at 300 mA/cm(exp 2) current density and an operating temperature of 90 C. A deterrent to performance appears to be methanol crossover through the membrane to the oxygen electrode. Further improvements in performance appear possible by minimizing the methanol crossover rate.

  11. Advanced supersonic technology fuel tank sealants

    NASA Technical Reports Server (NTRS)

    Rosser, R. W.; Parker, J. A.

    1976-01-01

    Status of the fuel tank simulation and YF-12A flight tests utilizing a fluorosilicone sealant is described. New elastomer sealant development is detailed, and comparisons of high and low temperature characteristics are made to baseline fluorosilicone sealants.

  12. Materials/manufacturing element of the Advanced Turbine System Program

    SciTech Connect

    Karnitz, M.A.; Devan, J.H.; Holcomb, R.S.; Ferber, M.K.; Harrison, R.W.

    1994-08-01

    One of the supporting elements of the Advanced Turbine Systems (ATS) Program is the materials/manufacturing technologies task. The objective of this element is to address critical materials issues for both industrial and utility gas turbines. DOE Oak Ridge Operations Office (ORO) will manage this element of the program, and a team from DOE-ORO and Oak Ridge National Laboratory is coordinating the planning for the materials/manufacturing effort. This paper describes that planning activity which is in the early stages.

  13. Development of advanced fuel cell system, phase 2

    NASA Technical Reports Server (NTRS)

    Handley, L. M.; Meyer, A. P.; Bell, W. F.

    1973-01-01

    A multiple task research and development program was performed to improve the weight, life, and performance characteristics of hydrogen-oxygen alkaline fuel cells for advanced power systems. Development and characterization of a very stable gold alloy catalyst was continued from Phase I of the program. A polymer material for fabrication of cell structural components was identified and its long term compatibility with the fuel cell environment was demonstrated in cell tests. Full scale partial cell stacks, with advanced design closed cycle evaporative coolers, were tested. The characteristics demonstrated in these tests verified the feasibility of developing the engineering model system concept into an advanced lightweight long life powerplant.

  14. 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.

  15. 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.

  16. 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.

  17. Fuel Properties Database from the Alternative Fuels and Advanced Vehicles Data Center (AFDC)

    DOE Data Explorer

    This database contains information on advanced petroleum and non-petroleum based fuels, as well as key data on advanced compression ignition fuels. Included are data on physical, chemical, operational, environmental, safety, and health properties. These data result from tests conducted according to standard methods (mostly American Society for Testing and Materials (ASTM). The source and test methods for each fuel data set are provided with the information. The database can be searched in various ways and can output numbers or explanatory text. Heavy vehicle chassis emission data are also available for some fuels.

  18. A fuel conservation study for transport aircraft utilizing advanced technology and hydrogen fuel

    NASA Technical Reports Server (NTRS)

    Berry, W.; Calleson, R.; Espil, J.; Quartero, C.; Swanson, E.

    1972-01-01

    The conservation of fossil fuels in commercial aviation was investigated. Four categories of aircraft were selected for investigation: (1) conventional, medium range, low take-off gross weight; (2) conventional, long range, high take-off gross weights; (3) large take-off gross weight aircraft that might find future applications using both conventional and advanced technology; and (4) advanced technology aircraft of the future powered with liquid hydrogen fuel. It is concluded that the hydrogen fueled aircraft can perform at reduced size and gross weight the same payload/range mission as conventionally fueled aircraft.

  19. 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)

  20. 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.

  1. Surrogate Model Development for Fuels for Advanced Combustion Engines

    SciTech Connect

    Anand, Krishnasamy; Ra, youngchul; Reitz, Rolf; Bunting, Bruce G

    2011-01-01

    The fuels used in internal-combustion engines are complex mixtures of a multitude of different types of hydrocarbon species. Attempting numerical simulations of combustion of real fuels with all of the hydrocarbon species included is highly unrealistic. Thus, a surrogate model approach is generally adopted, which involves choosing a few representative hydrocarbon species whose overall behavior mimics the characteristics of the target fuel. The present study proposes surrogate models for the nine fuels for advanced combustion engines (FACE) that have been developed for studying low-emission, high-efficiency advanced diesel engine concepts. The surrogate compositions for the fuels are arrived at by simulating their distillation profiles to within a maximum absolute error of 4% using a discrete multi-component (DMC) fuel model that has been incorporated in the multi-dimensional computational fluid dynamics (CFD) code, KIVA-ERC-CHEMKIN. The simulated surrogate compositions cover the range and measured concentrations of the various hydrocarbon classes present in the fuels. The fidelity of the surrogate fuel models is judged on the basis of matching their specific gravity, lower heating value, hydrogen/carbon (H/C) ratio, cetane number, and cetane index with the measured data for all nine FACE fuels.

  2. Advanced composite polymer electrolyte fuel cell membranes

    SciTech Connect

    Wilson, M.S.; Zawodzinski, T.A.; Gottesfeld, S.; Kolde, J.A.; Bahar, B.

    1995-09-01

    A new type of reinforced composite perfluorinated polymer electrolyte membrane, GORE-SELECT{trademark} (W.L. Gore & Assoc.), is characterized and tested for fuel cell applications. Very thin membranes (5-20 {mu}m thick) are available. The combination of reinforcement and thinness provides high membrane, conductances (80 S/cm{sup 2} for a 12 {mu}m thick membrane at 25{degrees}C) and improved water distribution in the operating fuel cell without sacrificing longevity or durability. In contrast to nonreinforced perfluorinated membranes, the x-y dimensions of the GORE-SELECT membranes are relatively unaffected by the hydration state. This feature may be important from the viewpoints of membrane/electrode interface stability and fuel cell manufacturability.

  3. Fuel cell and advanced turbine power cycle

    SciTech Connect

    White, D.J.

    1995-10-19

    Solar Turbines, Incorporated (Solar) has a vested interest in the integration of gas turbines and high temperature fuel cells and in particular, solid oxide fuel cells (SOFCs). Solar has identified a parallel path approach to the technology developments needed for future products. The primary approach is to move away from the simple cycle industrial machines of the past and develop as a first step more efficient recuperated engines. This move was prompted by the recognition that the simple cycle machines were rapidly approaching their efficiency limits. Improving the efficiency of simple cycle machines is and will become increasingly more costly. Each efficiency increment will be progressively more costly than the previous step.

  4. Advanced technology lightweight fuel cell program

    NASA Technical Reports Server (NTRS)

    Martin, R. E.

    1981-01-01

    The potential of the alkaline electrolyte fuel cell as the power source in a multi hundred kilowatt orbital energy storage system was studied. The total system weight of an electrolysis cell energy storage system was determined. The tests demonstrated: (1) the performance stability of a platinum on carbon anode catalyst configuration after 5000 hours of testing has no loss in performance; (2) capability of the alkaline fuel cell to operate to a cyclical load profile; (3) suitability of a lightweight graphite electrolyte reservoir plate for use in the alkaline fuel cell; (4) long life potential of a hybrid polysulfone cell edge frame construction; and (5) long term stability of a fiber reinforced potassium titanate matrix structure. The power section tested operates with passive water removal eliminating the requirement for a dynamic hydrogen pump water separator thereby allowing a powerplant design with reduced weight, lower parasite power, and a potential for high reliability and extended endurance. It is concluded that two perovskites are unsuitable for use as a catalyst or as a catalyst support at the cathode of an alkaline fuel cell.

  5. 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.

  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. Ultraclean Fuels Production and Utilization for the Twenty-First Century: Advances toward Sustainable Transportation Fuels

    SciTech Connect

    Fox, Elise B.; Liu, Zhong-Wen; Liu, Zhao-Tie

    2013-11-21

    Ultraclean fuels production has become increasingly important as a method to help decrease emissions and allow the introduction of alternative feed stocks for transportation fuels. Established methods, such as Fischer-Tropsch, have seen a resurgence of interest as natural gas prices drop and existing petroleum resources require more intensive clean-up and purification to meet stringent environmental standards. This review covers some of the advances in deep desulfurization, synthesis gas conversion into fuels and feed stocks that were presented at the 245th American Chemical Society Spring Annual Meeting in New Orleans, LA in the Division of Energy and Fuels symposium on "Ultraclean Fuels Production and Utilization".

  8. 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)

  9. 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

  10. Advanced Fuels Can Reduce the Cost of Getting Into Space

    NASA Technical Reports Server (NTRS)

    Palaszewski, Bryan A.

    1998-01-01

    Rocket propellant and propulsion technology improvements can reduce the development time and operational costs of new space vehicle programs, and advanced propellant technologies can make space vehicles safer and easier to operate, and can improve their performance. Five major areas have been identified for fruitful research: monopropellants, alternative hydrocarbons, gelled hydrogen, metallized gelled propellants, and high-energy-density propellants. During the development of the NASA Advanced Space Transportation Plan, these technologies were identified as those most likely to be effective for new NASA vehicles. Several NASA research programs had fostered work in fuels under the topic Fuels and Space Propellants for Reusable Launch Vehicles in 1996 to 1997. One component of this topic was to promote the development and commercialization of monopropellant rocket fuels, hypersonic fuels, and high-energy-density propellants. This research resulted in the teaming of small business with large industries, universities, and Government laboratories. This work is ongoing with seven contractors. The commercial products from these contracts will bolster advanced propellant research. Work also is continuing under other programs, which were recently realigned under the "Three Pillars" of NASA: Global Civil Aviation, Revolutionary Technology Leaps, and Access to Space. One of the five areas is described below, and its applications and effect on future missions is discussed. This work is being conducted at the NASA Lewis Research Center with the assistance of the NASA Marshall Space Flight Center. The regenerative cooling of spacecraft engines and other components can improve overall vehicle performance. Endothermic fuels can absorb energy from an engine nozzle and chamber and help to vaporize high-density fuel before it enters the combustion chamber. For supersonic and hypersonic aircraft, endothermic fuels can absorb the high heat fluxes created on the wing leading edges and

  11. 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)

  12. 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.

  13. Assessment for advanced fuel cycle options in CANDU

    SciTech Connect

    Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

    2013-07-01

    The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a driver fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.

  14. Recent advances in high-performance direct methanol fuel cells

    SciTech Connect

    Narayanan, S.R.; Chun, W.; Valdez, T.I.

    1996-12-31

    Direct methanol fuel cells for portable power applications have been advanced significantly under DARPA- and ARO-sponsored programs over the last five years. A liquid-feed direct methanol fuel cell developed under these programs, employs a proton exchange membrane as electrolyte and operates on aqueous solutions of methanol with air or oxygen as the oxidant. Power densities as high as 320 mW/cm{sup 2} have been demonstrated. Demonstration of five-cell stack based on the liquid-feed concept have been successfully performed by Giner Inc. and the Jet Propulsion Laboratory. Over 2000 hours of life-testing have been completed on these stacks. These fuel cells have been also been demonstrated by USC to operate on alternate fuels such as trimethoxymethane, dimethoxymethane and trioxane. Reduction in the parasitic loss of fuel across the fuel cell, a phenomenon termed as {open_quotes}fuel crossover{close_quotes} has been achieved using polymer membranes developed at USC. As a result efficiencies as high as 40% is considered attainable with this type of fuel cell. The state-of-development has reached a point where it is now been actively considered for stationary, portable and transportation applications. The research and development issues have been the subject of several previous articles and the present article is an attempt to summarize the key advances in this technology.

  15. Natural Gas for Advanced Dual-Fuel Combustion Strategies

    NASA Astrophysics Data System (ADS)

    Walker, Nicholas Ryan

    Natural gas fuels represent the next evolution of low-carbon energy feedstocks powering human activity worldwide. The internal combustion engine, the energy conversion device widely used by society for more than one century, is capable of utilizing advanced combustion strategies in pursuit of ultra-high efficiency and ultra-low emissions. Yet many emerging advanced combustion strategies depend upon traditional petroleum-based fuels for their operation. In this research the use of natural gas, namely methane, is applied to both conventional and advanced dual-fuel combustion strategies. In the first part of this work both computational and experimental studies are undertaken to examine the viability of utilizing methane as the premixed low reactivity fuel in reactivity controlled compression ignition, a leading advanced dual-fuel combustion strategy. As a result, methane is shown to be capable of significantly extending the load limits for dual-fuel reactivity controlled compression ignition in both light- and heavy-duty engines. In the second part of this work heavy-duty single-cylinder engine experiments are performed to research the performance of both conventional dual-fuel (diesel pilot ignition) and advanced dual-fuel (reactivity controlled compression ignition) combustion strategies using methane as the premixed low reactivity fuel. Both strategies are strongly influenced by equivalence ratio; diesel pilot ignition offers best performance at higher equivalence ratios and higher premixed methane ratios, whereas reactivity controlled compression ignition offers superior performance at lower equivalence ratios and lower premixed methane ratios. In the third part of this work experiments are performed in order to determine the dominant mode of heat release for both dual-fuel combustion strategies. By studying the dual-fuel homogeneous charge compression ignition and single-fuel spark ignition, strategies representative of autoignition and flame propagation

  16. Current Comparison of Advanced Nuclear Fuel Cycles

    SciTech Connect

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

    2007-04-01

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

  17. Engineering development of advanced physical fine coal cleaning for premium fuel applications

    SciTech Connect

    Smit, F.J.; Jha, M.C.

    1993-01-18

    This project is a step in the Department of Energy's program to show that ultra-clean fuel can be produced from selected coals and that the fuel will be a cost-effective replacement for oil and natural gas now fueling boilers in this country. The replacement of premium fossil fuels with coal can only be realized if retrofit costs are kept to a minimum and retrofit boiler emissions meet national goals for clean air. These concerns establish the specifications for maximum ash and sulfur levels and combustion properties of the ultra-clean coal. The primary objective is to develop the design base for prototype commercial advanced fine coal cleaning facilities capable of producing ultra-clean coals suitable for conversion to coal-water slurry fuel. The fine coal cleaning technologies are advanced column flotation and selective agglomeration. A secondary objective is to develop the design base for near-term commercial integration of advanced fine coal cleaning technologies in new or existing coal preparation plants for economically and efficiently processing minus 28-mesh coal fines. A third objective is to determine the distribution of toxic trace elements between clean coal and refuse when applying the advance column flotation and selective agglomeration technologies. The project team consists of Amax Research Development Center (Amax R D), Amax Coal industries, Bechtel Corporation, Center for Applied Energy Research (CAER) at the University of Kentucky, and Arcanum Corporation.

  18. Development of Advanced Fuel Cell System (Phase 4)

    NASA Technical Reports Server (NTRS)

    Meyer, A. P.; Bell, W. F.

    1976-01-01

    A multiple-task research and development program was performed to improve the weight, life, and performance characteristics of hydrogen-oxygen alkaline fuel cells for advanced power systems. During Phase 4, the lowest stabilized degradation rate observed in all the testing completed during four phases of the program, 1 microvolt/hour, was demonstrated. This test continues after 5,000 hours of operation. The cell incorporates a PPf anode, a 90Au/10Pt cathode, a hybrid frame, and a Fybex matrix. These elements were developed under this program to extend cell life. The result demonstrated that the 80Au/20Pt cathode is as stable as a 90Au/10Pt cathode of twice the precious metal loading, was confirmed in full-scale cells. A hybrid frame two-cell plaque with dedicated flow fields and manifolds for all fluids was demonstrated to prevent the cell-to cell electrolyte transfer that limited the endurance of multicell plaques. At the conclusion of Phase 4, more than 90,900 hours of testing had been completed and twelve different cell designs had been evaluated. A technology base has been established which is ready for evaluation at the powerplant level.

  19. Alternative Fuel and Advanced Technology Commercial Lawn Equipment (Brochure)

    SciTech Connect

    Not Available

    2014-10-01

    The U.S. Department of Energy's Clean Cities program produced this guide to help inform the commercial mowing industry about product options and potential benefits. This guide provides information about equipment powered by propane, ethanol, compressed natural gas, biodiesel, and electricity, as well as advanced engine technology. In addition to providing an overview for organizations considering alternative fuel lawn equipment, this guide may also be helpful for organizations that want to consider using additional alternative fueled equipment.

  20. ENHANCING ADVANCED CANDU PROLIFERATION RESISTANCE FUEL WITH MINOR ACTINIDES

    SciTech Connect

    Gray S. Chang

    2010-05-01

    The advanced nuclear system will significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. Minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

  1. Alternative Fuel and Advanced Technology Commercial Lawn Equipment

    SciTech Connect

    2014-10-10

    The U.S. Department of Energy's Clean Cities program produced this guide to help inform the commercial mowing industry about product options and potential benefits. This guide provides information about equipment powered by propane, ethanol, compressed natural gas, biodiesel, and electricity, as well as advanced engine technology. In addition to providing an overview for organizations considering alternative fuel lawn equipment, this guide may also be helpful for organizations that want to consider using additional alternative fueled equipment.

  2. Creep analysis of fuel plates for the Advanced Neutron Source

    SciTech Connect

    Swinson, W.F.; Yahr, G.T.

    1994-11-01

    The reactor for the planned Advanced Neutron Source will use closely spaced arrays of fuel plates. The plates are thin and will have a core containing enriched uranium silicide fuel clad in aluminum. The heat load caused by the nuclear reactions within the fuel plates will be removed by flowing high-velocity heavy water through narrow channels between the plates. However, the plates will still be at elevated temperatures while in service, and the potential for excessive plate deformation because of creep must be considered. An analysis to include creep for deformation and stresses because of temperature over a given time span has been performed and is reported herein.

  3. IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

    SciTech Connect

    I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

    2010-10-01

    Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

  4. Advanced space power PEM fuel cell systems

    NASA Technical Reports Server (NTRS)

    Vanderborgh, N. E.; Hedstrom, J.; Huff, J. R.

    1989-01-01

    A model showing mass and heat transfer in proton exchange membrane (PEM) single cells is presented. For space applications, stack operation requiring combined water and thermal management is needed. Advanced hardware designs able to combine these two techniques are available. Test results are shown for membrane materials which can operate with sufficiently fast diffusive water transport to sustain current densities of 300 ma per square centimeter. Higher power density levels are predicted to require active water removal.

  5. Optical Fuel Injector Patternation Measurements in Advanced Liquid-Fueled, High Pressure, Gas Turbine Combustors

    NASA Technical Reports Server (NTRS)

    Locke, R. J.; Hicks, Y. R.; Anderson, R. C.; Zaller, M. M.

    1998-01-01

    Planar laser-induced fluorescence (PLIF) imaging and planar Mie scattering are used to examine the fuel distribution pattern (patternation) for advanced fuel injector concepts in kerosene burning, high pressure gas turbine combustors. Three fuel injector concepts for aerospace applications were investigated under a broad range of operating conditions. Fuel PLIF patternation results are contrasted with those obtained by planar Mie scattering. For one injector, further comparison is also made with data obtained through phase Doppler measurements. Differences in spray patterns for diverse conditions and fuel injector configurations are readily discernible. An examination of the data has shown that a direct determination of the fuel spray angle at realistic conditions is also possible. The results obtained in this study demonstrate the applicability and usefulness of these nonintrusive optical techniques for investigating fuel spray patternation under actual combustor conditions.

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

    SciTech Connect

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

    2013-07-01

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

  7. Fuels for Advanced Combustion Engines Research Diesel Fuels: Analysis of Physical and Chemical Properties

    SciTech Connect

    Gallant, Tom; Franz, Jim; Alnajjar, Mikhail; Storey, John Morse; Lewis Sr, Samuel Arthur; Sluder, Scott; Cannella, William C; Fairbridge, Craig; Hager, Darcy; Dettman, Heather; Luecke, Jon; Ratcliff, Matthew A.; Zigler, Brad

    2009-01-01

    The CRC Fuels for Advanced Combustion Engines working group has worked to identify a matrix of research diesel fuels for use in advanced combustion research applications. Nine fuels were specified and formulated to investigate the effects of cetane number aromatic content and 90% distillation fraction. Standard ASTM analyses were performed on the fuels as well as GC/MS and /u1H//u1/u3C NMR analyses and thermodynamic characterizations. Details of the actual results of the fuel formulations compared with the design values are presented, as well as results from standard analyses, such as heating value, viscosity and density. Cetane number characterizations were accomplished by using both the engine method and the Ignition Quality Tester (IQT/sT) apparatus.

  8. Advanced coal-fueled gas turbine systems reference system definition update

    SciTech Connect

    Not Available

    1991-09-01

    The objective of the the Direct Coal-Fueled 80 MW Combustion Turbine Program is to establish the technology required for private sector use of an advanced coal-fueled combustion turbine power system. Under this program the technology for a direct coal-fueled 80 MW combustion turbine is to be developed. This unit would be an element in a 207 MW direct coal-fueled combustion turbine combined cycle which includes two combustion turbines, two heat recovery steam generators and a steam turbine. Key to meeting the program objectives is the development of a successful high pressure slagging combustor that burns coal, while removing sulfur, particulates, and corrosive alkali matter from the combustion products. Westinghouse and Textron (formerly AVCO Research Laboratory/Textron) have designed and fabricated a subscale slagging combustor. This slagging combustor, under test since September 1988, has been yielding important experimental data, while having undergone several design iterations.

  9. Spherically-Convergent, Advanced-Fuel Systems

    NASA Astrophysics Data System (ADS)

    Barnes, D. C.; Nebel, R. A.; Schauer, M. M.; Umstadter, K. R.

    1998-11-01

    Combining nonneutral electron confinement with spherical ion convergence leads to a cm sized reactor volume with high power density.(R. A. Nebel and D. C. Barnes, Fusion Technol.), to appear (1998); D. C. Barnes and R. A. Nebel, Phys. of Plasmas 5, 2498 (1998). This concept is being investigated experimentally,(D. C. Barnes, T. B. Mitchell, and M. M. Schauer, Phys. Plasmas) 4, 1745 (1997). and results will be reported. We argue that D-D operation of such a system offers all the advantages of aneutronic fusion cycles. In particular, no breeding or large tritium inventory is required, and material problems seem tractable based on previous LWR experience. In addition the extremely small unit size leads to a massively modular system which is easily maintained and repaired, suggesting a very high availability. It may also be possible to operate such a system with low or aneutronic fuels. Preliminary work in this direction will be presented.

  10. 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.

  11. Component Development - Advanced Fuel Cells for Transportation Applications

    SciTech Connect

    Butler, William

    2000-06-19

    Report summarizes results of second phase of development of Vairex air compressor/expander for automotive fuel cell power systems. Project included optimizing key system performance parameters, as well as reducing number of components and the project cost, size and weight of the air system. Objectives were attained. Advanced prototypes are in commercial test environments.

  12. Fuel-cycle greenhouse gas emissions impacts of alternative transportation fuels and advanced vehicle technologies.

    SciTech Connect

    Wang, M. Q.

    1998-12-16

    At an international conference on global warming, held in Kyoto, Japan, in December 1997, the United States committed to reduce its greenhouse gas (GHG) emissions by 7% over its 1990 level by the year 2012. To help achieve that goal, transportation GHG emissions need to be reduced. Using Argonne's fuel-cycle model, I estimated GHG emissions reduction potentials of various near- and long-term transportation technologies. The estimated per-mile GHG emissions results show that alternative transportation fuels and advanced vehicle technologies can help significantly reduce transportation GHG emissions. Of the near-term technologies evaluated in this study, electric vehicles; hybrid electric vehicles; compression-ignition, direct-injection vehicles; and E85 flexible fuel vehicles can reduce fuel-cycle GHG emissions by more than 25%, on the fuel-cycle basis. Electric vehicles powered by electricity generated primarily from nuclear and renewable sources can reduce GHG emissions by 80%. Other alternative fuels, such as compressed natural gas and liquefied petroleum gas, offer limited, but positive, GHG emission reduction benefits. Among the long-term technologies evaluated in this study, conventional spark ignition and compression ignition engines powered by alternative fuels and gasoline- and diesel-powered advanced vehicles can reduce GHG emissions by 10% to 30%. Ethanol dedicated vehicles, electric vehicles, hybrid electric vehicles, and fuel-cell vehicles can reduce GHG emissions by over 40%. Spark ignition engines and fuel-cell vehicles powered by cellulosic ethanol and solar hydrogen (for fuel-cell vehicles only) can reduce GHG emissions by over 80%. In conclusion, both near- and long-term alternative fuels and advanced transportation technologies can play a role in reducing the United States GHG emissions.

  13. Intercode Advanced Fuels and Cladding Comparison Using BISON, FRAPCON, and FEMAXI Fuel Performance Codes

    NASA Astrophysics Data System (ADS)

    Rice, Aaren

    As part of the Department of Energy's Accident Tolerant Fuels (ATF) campaign, new cladding designs and fuel types are being studied in order to help make nuclear energy a safer and more affordable source for power. This study focuses on the implementation and analysis of the SiC cladding and UN, UC, and U3Si2 fuels into three specific nuclear fuel performance codes: BISON, FRAPCON, and FEMAXI. These fuels boast a higher thermal conductivity and uranium density than traditional UO2 fuel which could help lead to longer times in a reactor environment. The SiC cladding has been studied for its reduced production of hydrogen gas during an accident scenario, however the SiC cladding is a known brittle and unyielding material that may fracture during PCMI (Pellet Cladding Mechanical Interaction). This work focuses on steady-state operation with advanced fuel and cladding combinations. By implementing and performing analysis work with these materials, it is possible to better understand some of the mechanical interactions that could be seen as limiting factors. In addition to the analysis of the materials themselves, a further analysis is done on the effects of using a fuel creep model in combination with the SiC cladding. While fuel creep is commonly ignored in the traditional UO2 fuel and Zircaloy cladding systems, fuel creep can be a significant factor in PCMI with SiC.

  14. Advanced coal-fueled gas turbine systems

    SciTech Connect

    Not Available

    1992-04-24

    No combustion tests for this program were conducted during this reporting period of January 1 to March 31, 1992. DOE-sponsored slogging combustor tests have been suspended since December 1991 in order to perform combustion tests on Northern States Power Company (NSP) coals. The NSP coal tests were conducted to evaluate combustor performance when burning western sub bituminous coals. The results of these tests will guide commercialization efforts, which are being promoted by NSP, Westinghouse Electric, and Textron Defense Systems. The NSP testing has been completed and preparation of the final report for that effort is underway. Although the NSP testing program has been completed, the Westinghouse/DOE program will not be resumed immediately. The reason for this is that Textron Defense Systems (TDS) has embarked on an internally funded program requiring installation of a new liquid fuel combustor system at the Haverhill site. The facility modifications for this new system are significant and it is not possible to continue the Westinghouse/DOE testing while these modifications are being made. These facility modifications are being performed during the period February 15, 1992 through May 31, 1992. The Westinghouse/DOE program can be resumed upon completion of this work.

  15. Advanced coal-fueled gas turbine systems

    NASA Astrophysics Data System (ADS)

    1992-04-01

    No combustion tests for this program were conducted during this reporting period of January 1 to March 31, 1992. DOE-sponsored slogging combustor tests have been suspended since December 1991 in order to perform combustion tests on Northern States Power Company (NSP) coals. The NSP coal tests were conducted to evaluate combustor performance when burning western sub bituminous coals. The results of these tests will guide commercialization efforts, which are being promoted by NSP, Westinghouse Electric, and Textron Defense Systems. The NSP testing has been completed and preparation of the final report for that effort is underway. Although the NSP testing program has been completed, the Westinghouse/DOE program will not be resumed immediately. The reason for this is that Textron Defense Systems (TDS) has embarked on an internally funded program requiring installation of a new liquid fuel combustor system at the Haverhill site. The facility modifications for this new system are significant and it is not possible to continue the Westinghouse/DOE testing while these modifications are being made. These facility modifications are being performed during the period February 15, 1992 through May 31, 1992. The Westinghouse/DOE program can be resumed upon completion of this work.

  16. Advanced materials for solid oxide fuel cells

    SciTech Connect

    Armstrong, T.R.; Stevenson, J.

    1995-08-01

    The purpose of this research is to improve the properties of the current state-of-the-art materials used for solid oxide fuel cells (SOFCs). The objectives are to: (1) develop materials based on modifications of the state-of-the-art materials; (2) minimize or eliminate stability problems in the cathode, anode, and interconnect; (3) Electrochemically evaluate (in reproducible and controlled laboratory tests) the current state-of-the-art air electrode materials and cathode/electrolyte interfacial properties; (4) Develop accelerated electrochemical test methods to evaluate the performance of SOFCs under controlled and reproducible conditions; and (5) Develop and test materials for use in low-temperature SOFCs. The goal is to modify and improve the current state-of-the-art materials and minimize the total number of cations in each material to avoid negative effects on the materials properties. Materials to reduce potential deleterious interactions, (3) improve thermal, electrical, and electrochemical properties, (4) develop methods to synthesize both state-of-the-art and alternative materials for the simultaneous fabricatoin and consolidation in air of the interconnections and electrodes with the solid electrolyte, and (5) understand electrochemical reactions at materials interfaces and the effects of component composition and processing on those reactions.

  17. Fuel savings potential of the NASA Advanced Turboprop Program

    NASA Technical Reports Server (NTRS)

    Whitlow, J. B., Jr.; Sievers, G. K.

    1984-01-01

    The NASA Advanced Turboprop (ATP) Program is directed at developing new technology for highly loaded, multibladed propellers for use at Mach 0.65 to 0.85 and at altitudes compatible with the air transport system requirements. Advanced turboprop engines offer the potential of 15 to 30 percent savings in aircraft block fuel relative to advanced turbofan engines (50 to 60 percent savings over today's turbofan fleet). The concept, propulsive efficiency gains, block fuel savings and other benefits, and the program objectives through a systems approach are described. Current program status and major accomplishments in both single rotation and counter rotation propeller technology are addressed. The overall program from scale model wind tunnel tests to large scale flight tests on testbed aircraft is discussed.

  18. 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.

  19. Assessment of Research Needs for Advanced Fuel Cells

    SciTech Connect

    Penner, S.S.

    1985-11-01

    The DOE Advanced Fuel Cell Working Group (AFCWG) was formed and asked to perform a scientific evaluation of the current status of fuel cells, with emphasis on identification of long-range research that may have a significant impact on the practical utilization of fuel cells in a variety of applications. The AFCWG held six meetings at locations throughout the country where fuel cell research and development are in progress, for presentations by experts on the status of fuel cell research and development efforts, as well as for inputs on research needs. Subsequent discussions by the AFCWG have resulted in the identification of priority research areas that should be explored over the long term in order to advance the design and performance of fuel cells of all types. Surveys describing the salient features of individual fuel cell types are presented in Chapters 2 to 6 and include elaborations of long-term research needs relating to the expeditious introduction of improved fuel cells. The Introduction and the Summary (Chapter 1) were prepared by AFCWG. They were repeatedly revised in response to comments and criticism. The present version represents the closest approach to a consensus that we were able to reach, which should not be interpreted to mean that each member of AFCWG endorses every statement and every unexpressed deletion. The Introduction and Summary always represent a majority view and, occasionally, a unanimous judgment. Chapters 2 to 6 provide background information and carry the names of identified authors. The identified authors of Chapters 2 to 6, rather than AFCWG as a whole, bear full responsibility for the scientific and technical contents of these chapters.

  20. Steady-State Analysis Model for Advanced Fuel Cycle Schemes.

    Energy Science and Technology Software Center (ESTSC)

    2008-03-17

    Version 00 SMAFS was developed as a part of the study, "Advanced Fuel Cycles and Waste Management", which was performed during 2003-2005 by an ad-hoc expert group under the Nuclear Development Committee in the OECD/NEA. The model was designed for an efficient conduct of nuclear fuel cycle scheme cost analyses. It is simple, transparent and offers users the capability to track down cost analysis results. All the fuel cycle schemes considered in the model aremore » represented in a graphic format and all values related to a fuel cycle step are shown in the graphic interface, i.e., there are no hidden values embedded in the calculations. All data on the fuel cycle schemes considered in the study including mass flows, waste generation, cost data, and other data such as activities, decay heat and neutron sources of spent fuel and high-level waste along time are included in the model and can be displayed. The user can easily modify values of mass flows and/or cost parameters and see corresponding changes in the results. The model calculates: front-end fuel cycle mass flows such as requirements of enrichment and conversion services and natural uranium; mass of waste based on the waste generation parameters and the mass flow; and all costs.« less

  1. 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

  2. Elemental sulfur recovery from desulfurization sorbents in advanced power systems

    SciTech Connect

    Dorchak, T.P.; Gangwal, S.K.; Turk, B.S.

    1995-12-31

    Regenerable metal oxide sorbents, such as zinc titanate, are being developed to efficiently remove hydrogen sulfide (H{sub 2}S) from coal gas in advanced power systems. Dilute air regeneration of the sorbents produces a tailgas containing a few percent sulfur dioxide (SO{sub 2}). Catalytic reduction of the SO{sub 2} to elemental sulfur with a coal gas slipstream using the Direct Sulfur Recovery Process (DSRP) is a leading first-generation technology. Currently the DSRP is undergoing field testing at gasifier sites. The objective of this study is to develop second-generation processes that produce elemental sulfur with limited use of coal gas. Novel approaches that were evaluated to produce elemental sulfur from sulfided sorbents include (1) SO{sub 2} regeneration, (2) substoichiometric oxidation, (3) steam regeneration followed by H{sub 2}S oxidation, and (4) steam-air regeneration. Experimental results at high temperature and high pressure demonstrate that, with simple sorbent modifications, direct regeneration to elemental sulfur is feasible without the use of coal gas.

  3. 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

  4. 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/.

  5. Fuel conservation merits of advanced turboprop transport aircraft

    NASA Technical Reports Server (NTRS)

    Revell, J. D.; Tullis, R. H.

    1977-01-01

    The advantages of a propfan powered aircraft for the commercial air transportation system were assessed by the comparison with an equivalent turbofan transport. Comparisons were accomplished on the basis of fuel utilization and operating costs, as well as aircraft weight and size. Advantages of the propfan aircraft, concerning fuel utilization and operating costs, were accomplished by considering: (1) incorporation of propfan performance and acoustic data; (2) revised mission profiles (longer design range and reduction in; and cruise speed) (3) utilization of alternate and advanced technology engines.

  6. Advanced thermally stable jet fuels. Technical progress report, 1995

    SciTech Connect

    Schobert, H.H.; Eser, S.; Song, C.

    1996-04-01

    The Penn State program in advanced thermally stable jet fuels has five components:(1) development of mechanisms of degradation and solids formation; (2) quantitative measurement of growth of sub- micrometer and micrometer sized particles suspended in fuels during thermal stressing; (3) characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) elucidation of the role of additives in retarding the formation of carbonaceous solids; and (5) assessment of the potential of producing high yields of cycloalkanes and hydroaromatics by direct liquefaction of coal. Progress reports for these tasks are presented.

  7. Hydrogen-bromine fuel cell advance component development

    NASA Technical Reports Server (NTRS)

    Charleston, Joann; Reed, James

    1988-01-01

    Advanced cell component development is performed by NASA Lewis to achieve improved performance and longer life for the hydrogen-bromine fuel cells system. The state-of-the-art hydrogen-bromine system utilizes the solid polymer electrolyte (SPE) technology, similar to the SPE technology developed for the hydrogen-oxygen fuel cell system. These studies are directed at exploring the potential for this system by assessing and evaluating various types of materials for cell parts and electrode materials for Bromine-hydrogen bromine environment and fabricating experimental membrane/electrode-catalysts by chemical deposition.

  8. Advanced fuels for plutonium management in pressurized water reactors

    NASA Astrophysics Data System (ADS)

    Vasile, A.; Dufour, Ph; Golfier, H.; Grouiller, J. P.; Guillet, J. L.; Poinot, Ch; Youinou, G.; Zaetta, A.

    2003-06-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate.

  9. Structural thermal tests on Advanced Neutron Source reactor fuel plates

    SciTech Connect

    Swinson, W.F.; Battiste, R.L.; Yahr, G.T.

    1995-08-01

    The thin aluminum-clad fuel plates proposed for the Advanced Neutron Source reactor are stressed by the high-velocity coolant flowing on each side of the plates and by the thermal gradients in the plates. The total stress, composed of the sum of the flow stress and the thermal stress at a point, could be reduced if the thermal loads tend to relax when the stress magnitude approaches the yield stress of the material. The potential of this occurring would be very significant in assessing the structural reliability of the fuel plates and has been investigated through experiment. The results of this investigation are given in this report.

  10. Advanced technology for extended endurance alkaline fuel cells

    NASA Technical Reports Server (NTRS)

    Sheibley, D. W.; Martin, R. A.

    1987-01-01

    Advanced components have been developed for alkaline fuel cells with a view to the satisfaction of NASA Space Station design requirements for extended endurance. The components include a platinum-on-carbon catalyst anode, a potassium titanate-bonded electrolyte matrix, a lightweight graphite electrolyte reservoir plate, a gold-plated nickel-perforated foil electrode substrate, a polyphenylene sulfide cell edge frame material, and a nonmagnesium cooler concept. When incorporated into the alkaline fuel cell unit, these components are expected to yield regenerative operation in a low earth orbit Space Station with a design life greater than 5 years.

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

    SciTech Connect

    E. R. Johnson; R. E. Best

    2009-12-28

    materials were produced and consumed in a fleet of 100 1,000 MWe LWRs and in FRs. The model also included recycle and reuse of extant inventories of spent LWR fuel. The results of the simulations allowed comparisons of the two fuel cycles from the standpoints of cost, non-proliferation, radiological health, wastes generated, and sustainability. Results of the research also provide insights regarding (i) multiple recycling of uranium and plutonium from spent MOX fuel in LWRs, (ii) costs and benefits of reenriching and reusing uranium from spent LWR fuel; (iii) effects of using uranium, plutonium, and minor actinides from LWR spent fuels to produce fuel for FRs; (iv) net rates of consumption (burning) in FRs of actinide elements produced in LWRs, and (v) ependencies of and interactions among the different systems of an advanced nuclear fuel cycle -- and the flows of nuclear materials between these systems.

  12. On-Going Comparison of Advanced Fuel Cycle Options

    SciTech Connect

    Steven J. Piet; Ralph G. Bennett; Brent W. Dixon; J. Stephen Herring; David E. Shropshire; Mark Roth; J. D. Smith; Robert Hill; James Laidler; Kemal Pasamehmetoglu

    2004-10-01

    The Advanced Fuel Cycle Initiative (AFCI) program is addressing key issues associated with critical national needs. This paper compares the major options with these major “outcome” objectives - waste geological repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety as well as “process” objectives associated with readiness to proceed and adaptability and robustness in the face of uncertainties. Working together, separation, transmutation, and fuel technologies provide complete energy systems that can improve waste management compared to the current “once-through/no separation” approach. Future work will further increase confidence in potential solutions, optimize solutions for the mixtures of objectives, and develop attractive development and deployment paths for selected options. This will allow the nation to address nearer-term issues such as avoiding the need for additional geological repositories while making nuclear energy a more sustainable energy option for the long-term. While the Generation IV Initiative is exploring multiple reactor options for future nuclear energy for both electricity generation and additional applications, the AFCI is assessing fuel cycles options for either a continuation or expansion of nuclear energy in the United States. This report compares strategies and technology options for managing the associated spent fuel. There are four major potential strategies, as follows: · The current U.S. strategy is once through: standard nuclear power plants, standard fuel burnup, direct geological disposal of spent fuel. Variants include higher burnup fuels in water-cooled power plants, once-through gas-cooled power plants, and separation (without recycling) of spent fuel to reduce the number and cost of geological waste packages. · The second strategy is thermal recycle, recycling some fuel components in thermal reactors. This strategy extends the useful life of

  13. 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

  14. Prospects for advanced coal-fuelled fuel cell power plants

    NASA Astrophysics Data System (ADS)

    Jansen, D.; Vanderlaag, P. C.; Oudhuis, A. B. J.; Ribberink, J. S.

    1994-04-01

    As part of ECN's in-house R&D programs on clean energy conversion systems with high efficiencies and low emissions, system assessment studies have been carried out on coal gasification power plants integrated with high-temperature fuel cells (IGFC). The studies also included the potential to reduce CO2 emissions, and to find possible ways for CO2 extraction and sequestration. The development of this new type of clean coal technology for large-scale power generation is still far off. A significant market share is not envisaged before the year 2015. To assess the future market potential of coal-fueled fuel cell power plants, the promise of this fuel cell technology was assessed against the performance and the development of current state-of-the-art large-scale power generation systems, namely the pulverized coal-fired power plants and the integrated coal gasification combined cycle (IGCC) power plants. With the anticipated progress in gas turbine and gas clean-up technology, coal-fueled fuel cell power plants will have to face severe competition from advanced IGCC power plants, despite their higher efficiency.

  15. Advanced coal-fueled industrial cogeneration gas turbine system

    SciTech Connect

    LeCren, R.T.; Cowell, L.H.; Galica, M.A.; Stephenson, M.D.; Wen, C.S.

    1991-07-01

    Advances in coal-fueled gas turbine technology over the past few years, together with recent DOE-METC sponsored studies, have served to provide new optimism that the problems demonstrated in the past can be economically resolved and that the coal-fueled gas turbine can ultimately be the preferred system in appropriate market application sectors. The objective of the Solar/METC program is to prove the technical, economic, and environmental feasibility of a coal-fired gas turbine for cogeneration applications through tests of a Centaur Type H engine system operated on coal fuel throughout the engine design operating range. The five-year program consists of three phases, namely: (1) system description; (2) component development; (3) prototype system verification. A successful conclusion to the program will initiate a continuation of the commercialization plan through extended field demonstration runs.

  16. Coating parameters of zirconium carbide on advanced TRISO fuels

    NASA Astrophysics Data System (ADS)

    Dulude, Michael C.

    The feasibility of using very high temperature reactors (VHTR) as part of the next generation of nuclear reactors greatly depends on the tri-structural isotropic (TRISO) fuel particles reliability to retain both gaseous and metallic fission products created in irradiated UO2. Most research devoted to TRISO fuel particles has focused on the characteristics and retention ability of silicon carbide as the main barrier against metallic fission products. This work investigates the deposition parameters necessary to create advanced TRISO particles consisting of the standard SiC TRISO coatings with an additional layer of ZrC applied directly to the UO2 fuel kernel. The additional ZrC layer will act as an oxygen getter to prevent failure mechanisms experienced in TRISO particles. Two failure mechanisms that are of the most concern are the over pressurization of the particles and kernel migration within the TRISO particles. In this study successful ZrC coatings were created and the deposition characteristics were analyzed via optical and SEM microscopy techniques. The ZrC layer was confirmed through XRD analysis. This investigation also reduced U3O8 microspheres to UO2 in an argon atmosphere. The oxygen to metal ratio from the reduced U3O8 was back calculated from oxidation analysis performed with a TGA machine. Once consistent repeatability is shown with coating surrogate zirconia kernels, advanced TRISO coatings will be deposited on the UO2 fuel kernels.

  17. Systematic analysis of advanced fusion fuel in inertial fusion energy

    NASA Astrophysics Data System (ADS)

    Velarde, G.; Eliezer, S.; Henis, Z.; Piera, M.; Martinez-Val, J. M.

    1997-04-01

    Aneutronic fusion reactions can be considered as the cleanest way to exploit nuclear energy. However, these reactions present in general two main drawbacks.—very high temperatures are needed to reach relevant values of their cross sections—Moderate (and even low) energy yield per reaction. This value is still lower if measured in relation to the Z number of the reacting particles. It is already known that bremsstrahlung overruns the plasma reheating by fusion born charged-particles in most of the advanced fuels. This is for instance the case for proton-boron-11 fusion in a stoichiometric plasma and is also so in lithium isotopes fusion reactions. In this paper, the use of deuterium-tritium seeding is suggested to allow to reach higher burnup fractions of advanced fuels, starting at a lower ignition temperature. Of course, neutron production increases as DT contents does. Nevertheless, the ratio of neutron production to energy generation is much lower in DT-advanced fuel mixtures than in pure DT plasmas. One of the main findings of this work is that some natural resources (as D and Li-7) can be burned-up in a catalytic regime for tritium. In this case, neither external tritium breeding nor tritium storage are needed, because the tritium inventory after the fusion burst is the same as before it. The fusion reactor can thus operate on a pure recycling of a small tritium inventory.

  18. Thermal Behavior of Advanced UO{sub 2} Fuel at High Burnup

    SciTech Connect

    Muller, E.; Lambert, T.; Silberstein, K.; Therache, B.

    2007-07-01

    To improve the fuel performance, advanced UO{sub 2} products are developed to reduce significantly Pellet-Cladding Interaction and Fission Gas Release to increase high burnup safety margins on Light Water Reactors. To achieve the expected improvements, doping elements are currently used, to produce large grain viscoplastic UO{sub 2} fuel microstructures. In that scope, AREVA NP is conducting the qualification of a new UO{sub 2} fuel pellet obtained by optimum chromium oxide doping. To assess the fuel thermal performance, especially the fuel conductivity degradation with increasing burnup and also the kinetics of fission gas release under transient operating conditions, an instrumented in-pile experiment, called REMORA, has been developed by the CEA. One segment base irradiated for five cycles in a French EDF commercial PWR ({approx} 62 GWd/tM) was consequently re-instrumented with a fuel centerline thermocouple and an advanced pressure sensor. The design of this specific sensor is based on the counter-pressure principle and avoids any drift phenomenon due to nuclear irradiation. This rodlet was then irradiated in the GRIFFONOS rig of the Osiris experimental reactor at CEA Saclay. This device, located in the periphery of the core, is designed to perform test under conditions close to those prevailing in French PWR reactor. Power variations are carried out by translating the device relatively to the core. Self - powered neutron detectors are positioned in the loop in order to monitor the power the whole time of the irradiation. The re-irradiation of the REMORA experiment consisted of a stepped ramp to power in order to point out a potential degradation of the fuel thermal conductivity with increasing burnup. During the first part of the irradiation, most of the measurements were performed at low power in order to take into account the irradiation effects on UO{sub 2} thermal conductivity at high burnup in low range of temperature. The second part of the irradiation

  19. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  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. Combustion behaviors of a compression-ignition engine fueled with diesel/methanol blends under various fuel delivery advance angles.

    PubMed

    Huang, Zuohua; Lu, Hongbing; Jiang, Deming; Zeng, Ke; Liu, Bing; Zhang, Junqiang; Wang, Xibin

    2004-12-01

    A stabilized diesel/methanol blend was described and the basic combustion behaviors based on the cylinder pressure analysis was conducted in a compression-ignition engine. The study showed that increasing methanol mass fraction of the diesel/methanol blends would increase the heat release rate in the premixed burning phase and shorten the combustion duration of the diffusive burning phase. The ignition delay increased with the advancing of the fuel delivery advance angle for both the diesel fuel and the diesel/methanol blends. For a specific fuel delivery advance angle, the ignition delay increased with the increase of the methanol mass fraction (oxygen mass fraction) in the fuel blends and the behaviors were more obvious at low engine load and/or high engine speed. The rapid burn duration and the total combustion duration increased with the advancing of the fuel delivery advance angle. The centre of the heat release curve was close to the top-dead-centre with the advancing of the fuel delivery advance angle. Maximum cylinder gas pressure increased with the advancing of the fuel delivery advance angle, and the maximum cylinder gas pressure of the diesel/methanol blends gave a higher value than that of the diesel fuel. The maximum mean gas temperature remained almost unchanged or had a slight increase with the advancing of the fuel delivery advance angle, and it only slightly increased for the diesel/methanol blends compared to that of the diesel fuel. The maximum rate of pressure rise and the maximum rate of heat release increased with the advancing of the fuel delivery advance angle of the diesel/methanol blends and the value was highest for the diesel/methanol blends. PMID:15288277

  3. 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.

  4. The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles

    SciTech Connect

    Finck, P.; Edelstein, N.; Allen, T.; Burns, C.; Chadwick, M.; Corradini, M.; Dixon, D.; Goff, M.; Laidler, J.; McCarthy, K.; Moyer, B.; Nash, K.; Navrotsky, A.; Oblozinsky, P.; Pasamehmetoglu, K.; Peterson, P.; Sackett, J.; Sickafus, K. E.; Tulenko, J.; Weber, W.; Morss, L.; Henry, G.

    2005-09-01

    The objective of this report is to identify new basic science that will be the foundation for advances in nuclear fuel-cycle technology in the near term, and for changing the nature of fuel cycles and of the nuclear energy industry in the long term. The goals are to enhance the development of nuclear energy, to maximize energy production in nuclear reactor parks, and to minimize radioactive wastes, other environmental impacts, and proliferation risks. The limitations of the once-through fuel cycle can be overcome by adopting a closed fuel cycle, in which the irradiated fuel is reprocessed and its components are separated into streams that are recycled into a reactor or disposed of in appropriate waste forms. The recycled fuel is irradiated in a reactor, where certain constituents are partially transmuted into heavier isotopes via neutron capture or into lighter isotopes via fission. Fast reactors are required to complete the transmutation of long-lived isotopes. Closed fuel cycles are encompassed by the Department of Energy?s Advanced Fuel Cycle Initiative (AFCI), to which basic scientific research can contribute. Two nuclear reactor system architectures can meet the AFCI objectives: a ?single-tier? system or a ?dual-tier? system. Both begin with light water reactors and incorporate fast reactors. The ?dual-tier? systems transmute some plutonium and neptunium in light water reactors and all remaining transuranic elements (TRUs) in a closed-cycle fast reactor. Basic science initiatives are needed in two broad areas: ? Near-term impacts that can enhance the development of either ?single-tier? or ?dual-tier? AFCI systems, primarily within the next 20 years, through basic research. Examples: Dissolution of spent fuel, separations of elements for TRU recycling and transmutation Design, synthesis, and testing of inert matrix nuclear fuels and non-oxide fuels Invention and development of accurate on-line monitoring systems for chemical and nuclear species in the nuclear

  5. Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing

    SciTech Connect

    Fletcher, James H.; Cox, Philip; Harrington, William J; Campbell, Joseph L

    2013-09-03

    ABSTRACT Project Title: Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing PROJECT OBJECTIVE The objective of the project was to advance portable fuel cell system technology towards the commercial targets of power density, energy density and lifetime. These targets were laid out in the DOE’s R&D roadmap to develop an advanced direct methanol fuel cell power supply that meets commercial entry requirements. Such a power supply will enable mobile computers to operate non-stop, unplugged from the wall power outlet, by using the high energy density of methanol fuel contained in a replaceable fuel cartridge. Specifically this project focused on balance-of-plant component integration and miniaturization, as well as extensive component, subassembly and integrated system durability and validation testing. This design has resulted in a pre-production power supply design and a prototype that meet the rigorous demands of consumer electronic applications. PROJECT TASKS The proposed work plan was designed to meet the project objectives, which corresponded directly with the objectives outlined in the Funding Opportunity Announcement: To engineer the fuel cell balance-of-plant and packaging to meet the needs of consumer electronic systems, specifically at power levels required for mobile computing. UNF used existing balance-of-plant component technologies developed under its current US Army CERDEC project, as well as a previous DOE project completed by PolyFuel, to further refine them to both miniaturize and integrate their functionality to increase the system power density and energy density. Benefits of UNF’s novel passive water recycling MEA (membrane electrode assembly) and the simplified system architecture it enabled formed the foundation of the design approach. The package design was hardened to address orientation independence, shock, vibration, and environmental requirements. Fuel cartridge and fuel subsystems were improved to ensure effective fuel

  6. ADVANCED HETEROGENEOUS REBURN FUEL FROM COAL AND HOG MANURE

    SciTech Connect

    Melanie D. Jensen; Ronald C. Timpe; Jason D. Laumb

    2003-09-01

    This study was performed to investigate whether the nitrogen content inherent in hog manure and alkali used as a catalyst during processing could be combined with coal to produce a reburn fuel that would result in advanced reburning NO{sub x} control without the addition of either alkali or ammonia/urea. Fresh hog manure was processed in a cold-charge, 1-gal, batch autoclave system at 275 C under a reducing atmosphere in the presence of an alkali catalyst. Instead of the expected organic liquid, the resulting product was a waxy solid material. The waxy nature of the material made size reduction and feeding difficult as the material agglomerated and tended to melt, plugging the feeder. The material was eventually broken up and sized manually and a water-cooled feeder was designed and fabricated. Two reburn tests were performed in a pilot-scale combustor. The first test evaluated a reburn fuel mixture comprising lignite and air-dried, raw hog manure. The second test evaluated a reburn fuel mixture made of lignite and the processed hog manure. Neither reburn fuel reduced NO{sub x} levels in the combustor flue gas. Increased slagging and ash deposition were observed during both reburn tests. The material-handling and ash-fouling issues encountered during this study indicate that the use of waste-based reburn fuels could pose practical difficulties in implementation on a larger scale.

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

    NASA Astrophysics Data System (ADS)

    Carlsen, Robert W.

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

  8. NREL - Advanced Vehicles and Fuels Basics - Center for Transportation Technologies and Systems 2010

    SciTech Connect

    2010-01-01

    We can improve the fuel economy of our cars, trucks, and buses by designing them to use the energy in fuels more efficiently. Researchers at the National Renewable Energy Laboratory (NREL) are helping the nation achieve these goals by developing transportation technologies like: advanced vehicle systems and components; alternative fuels; as well as fuel cells, hybrid electric, and plug-in hybrid vehicles. For a text version of this video visit http://www.nrel.gov/learning/advanced_vehicles_fuels.html

  9. Development of Kinetic Mechanisms for Next-Generation Fuels and CFD Simulation of Advanced Combustion Engines

    SciTech Connect

    Pitz, William J.; McNenly, Matt J.; Whitesides, Russell; Mehl, Marco; Killingsworth, Nick J.; Westbrook, Charles K.

    2015-12-17

    Predictive chemical kinetic models are needed to represent next-generation fuel components and their mixtures with conventional gasoline and diesel fuels. These kinetic models will allow the prediction of the effect of alternative fuel blends in CFD simulations of advanced spark-ignition and compression-ignition engines. Enabled by kinetic models, CFD simulations can be used to optimize fuel formulations for advanced combustion engines so that maximum engine efficiency, fossil fuel displacement goals, and low pollutant emission goals can be achieved.

  10. NREL - Advanced Vehicles and Fuels Basics - Center for Transportation Technologies and Systems 2010

    ScienceCinema

    None

    2013-05-29

    We can improve the fuel economy of our cars, trucks, and buses by designing them to use the energy in fuels more efficiently. Researchers at the National Renewable Energy Laboratory (NREL) are helping the nation achieve these goals by developing transportation technologies like: advanced vehicle systems and components; alternative fuels; as well as fuel cells, hybrid electric, and plug-in hybrid vehicles. For a text version of this video visit http://www.nrel.gov/learning/advanced_vehicles_fuels.html

  11. 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.

  12. Irradiation Test of Advanced PWR Fuel in Fuel Test Loop at HANARO

    SciTech Connect

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo; Park, Su Ki; Seo, Chul Gyo

    2007-07-01

    A new fuel test loop has been constructed in the research reactor HANARO at KAERI. The main objective of the FTL (Fuel Test Loop) is an irradiation test of a newly developed LWR fuel under PWR or Candu simulated conditions. The first test rod will be loaded within 2007 and its irradiation test will be continued until a rod average their of 62 MWd/kgU. A total of five test rods can be loaded into the IPS (In-Pile Section) and fuel centerline temperature, rod internal pressure and fuel stack elongation can be measured by an on-line real time system. A newly developed advanced PWR fuel which consists of a HANA{sup TM} alloy cladding and a large grain UO{sub 2} pellet was selected as the first test fuel in the FTL. The fuel cladding, the HANA{sup TM} alloy, is an Nb containing Zirconium alloy that has shown better corrosion and creep resistance properties than the current Zircaloy-4 cladding. A total of six types of HANA{sup TM} alloy were developed and two or three of these candidate alloys will be used as test rod cladding, which have shown a superior performance to the others. A large-grain UO{sub 2} pellet has a 14{approx}16 micron 2D diameter grain size for a reduction of a fission gas release at a high burnup. In this paper, characteristics of the FTL and IPS are introduced and the expected operation and irradiation conditions are summarized for the test periods. Also the preliminary fuel performance analysis results, such as the cladding oxide thickness, fission gas release and rod internal pressure, are evaluated from the test rod safety analysis aspects. (authors)

  13. Advanced fuel cell development. Progress report, October-December 1979

    SciTech Connect

    Pierce, R. D.; Kucera, G. H.; Kupperman, D. S.; Poeppel, R. B.; Sim, J. W.; Singh, R. N.; Smith, J. L.

    1980-05-01

    Advanced fuel cell research and development activities at Argonne National Laboratory (ANL) during the period October-December 1979 are described. These efforts have been directed toward understanding and improving components of molten carbonate fuel cells and have included operation of 10-cm square cells. The principal focus has been on the development of electrolyte structures (LiAlO/sub 2/ and Li/sub 2/CO/sub 3/-K/sub 2/CO/sub 3/) that have good electrolyte retention and mechanical properties as well as long-term stability. This effort included work on preparation of sintered LiAlO/sub 2/ as electrolyte support, use of a scanning laser acoustic microscope to evaluate electrolyte structures, and measurements of the thermal expansion coefficients of various mixtures of ..beta..-LiAlO/sub 2/ and carbonate eutectic.

  14. High efficiency fuel cell/advanced turbine power cycles

    SciTech Connect

    Morehead, H.

    1995-10-19

    An outline of the Westinghouse high-efficiency fuel cell/advanced turbine power cycle is presented. The following topics are discussed: The Westinghouse SOFC pilot manufacturing facility, cell scale-up plan, pressure effects on SOFC power and efficiency, sureCell versus conventional gas turbine plants, sureCell product line for distributed power applications, 20 MW pressurized-SOFC/gas turbine power plant, 10 MW SOFC/CT power plant, sureCell plant concept design requirements, and Westinghouse SOFC market entry.

  15. Recent advances in solid polymer electrolyte fuel cell technology

    SciTech Connect

    Ticianelli, E.A.; Srinivasan, S.; Gonzalez, E.R.

    1988-01-01

    With methods used to advance solid polymer electrolyte fuel cell technology, we are close to obtaining the goal of 1 A/cm/sup 2/ at 0.7. Higher power densities have been reported (2 A/cm/sup 2/ at 0.5 V) but only with high catalyst loading electrodes (2 mg/cm/sup 2/ and 4 mg/cm/sup 2/ at anode and cathode, respectively) and using a Dow membrane with a better conductivity and water retention characteristics. Work is in progress to ascertain performances of cells with Dow membrane impregnated electrodes and Dow membrane electrolytes. 5 refs., 6 figs.

  16. 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.

  17. Recent advances in direct methanol fuel cells at Los Alamos National Laboratory

    NASA Astrophysics Data System (ADS)

    Ren, Xiaoming; Zelenay, Piotr; Thomas, Sharon; Davey, John; Gottesfeld, Shimshon

    This paper describes recent advances in the science and technology of direct methanol fuel cells (DMFCs) made at Los Alamos National Laboratory (LANL). The effort on DMFCs at LANL includes work devoted to portable power applications, funded by the Defense Advanced Research Project Agency (DARPA), and work devoted to potential transport applications, funded by the US DOE. We describe recent results with a new type of DMFC stack hardware that allows to lower the pitch per cell to 2 mm while allowing low air flow and air pressure drops. Such stack technology lends itself to both portable power and potential transport applications. Power densities of 300 W/l and 1 kW/l seem achievable under conditions applicable to portable power and transport applications, respectively. DMFC power system analysis based on the performance of this stack, under conditions applying to transport applications (joint effort with U.C. Davis), has shown that, in terms of overall system efficiency and system packaging requirements, a power source for a passenger vehicle based on a DMFC could compete favorably with a hydrogen-fueled fuel cell system, as well as with fuel cell systems based on fuel processing on board. As part of more fundamental studies performed, we describe optimization of anode catalyst layers in terms of PtRu catalyst nature, loading and catalyst layer composition and structure. We specifically show that, optimized content of recast ionic conductor added to the catalyst layer is a sensitive function of the nature of the catalyst. Other elements of membrane/electrode assembly (MEA) optimization efforts are also described, highlighting our ability to resolve, to a large degree, a well-documented problem of polymer electrolyte DMFCs, namely "methanol crossover". This was achieved by appropriate cell design, enabling fuel utilization as high as 90% in highly performing DMFCs.

  18. 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.

  19. 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.

  20. Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.

    SciTech Connect

    Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage

  1. Application of a Plasma Mass Separator to Advanced LWR Spent Fuel Reprocessing

    SciTech Connect

    Freeman, Richard; Miller, Robert; Papay, Larry; Wagoner, John; Ahlfeld, Charles; Czerwinski, Ken

    2006-07-01

    The US Department of Energy (DOE) is investigating spent fuel reprocessing for the purposes of increasing the effective capacity of a deep geological repository, reducing the radiotoxicity of waste placed in the repository and conserving nuclear fuel resources. DOE is considering hydro-chemical processing of the spent fuel after cutting the fuel cladding and fuel dissolution in nitric acid. The front end process, known as UREX, is largely based on the PUREX process and extracts U, Tc as well as fission product gases. A number of additional processing steps have become known as UREX+. One of the steps includes a further chemical treatment of remove Cs and Sr to reduce repository heat load. Other steps include successive extraction of the actinides from residual fission products, including the lanthanides. The additional UREX+ processing renders the actinides suitable for burning as reactor fuel in an advanced reactor to convert actinides to shorter-lived fission products and to produce power. New methods for separating groups of elements by their atomic mass have been developed and can be exploited to enhance spent fuel reprocessing. These physical processes dry the waste streams so that they can be vaporized and singly ionized in plasma that is contained in longitudinal magnetic and perpendicular electric fields. Proper configuration of the fields causes the plasma to rapidly rotate and expel heavier mass ions at the center of the machine. Lower mass ions form closed orbits within the cylindrical plasma column and are transported to either end of the machine. This plasma mass separator was originally developed to reduce the mass of material that must be immobilized in borosilicate glass from DOE defense waste at former weapons production facilities. The plasma mass separator appears to be well-suited for processing the UREX raffinate and solids streams by exploiting the large atomic mass gap that exists between lanthanides (< {approx}180 amu) and actinides

  2. 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.

  3. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  4. 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.

  5. 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)

  6. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    SciTech Connect

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  7. THE MISSION AND ACCOMPLISHMENTS FROM DOE’S FUEL CYCLE RESEARCH AND DEVELOPMENT (FCRD) ADVANCED FUELS CAMPAIGN

    SciTech Connect

    J. Carmack; L. Braase; F. Goldner

    2015-09-01

    The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors, enhance proliferation resistance of nuclear fuel, effectively utilize nuclear energy resources, and address the longer-term waste management challenges. This includes development of a state of the art Research and Development (R&D) infrastructure to support the use of a “goal oriented science based approach.” AFC uses a “goal oriented, science based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. One of the most challenging aspects of AFC is the management, integration, and coordination of major R&D activities across multiple organizations. AFC interfaces and collaborates with Fuel Cycle Technologies (FCT) campaigns, universities, industry, various DOE programs and laboratories, federal agencies (e.g., Nuclear Regulatory Commission [NRC]), and international organizations. Key challenges are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Challenged with the research and development of fuels for two different reactor technology platforms, AFC targeted transmutation fuel development and focused ceramic fuel development for Advanced LWR Fuels.

  8. Advances in Architectural Elements For Future Missions to Titan

    NASA Astrophysics Data System (ADS)

    Reh, Kim; Coustenis, Athena; Lunine, Jonathan; Matson, Dennis; Lebreton, Jean-Pierre; Vargas, Andre; Beauchamp, Pat; Spilker, Tom; Strange, Nathan; Elliott, John

    2010-05-01

    The future exploration of Titan is of high priority for the solar system exploration community as recommended by the 2003 National Research Council (NRC) Decadal Survey [1] and ESA's Cosmic Vision Program themes. Recent Cassini-Huygens discoveries continue to emphasize that Titan is a complex world with very many Earth-like features. Titan has a dense, nitrogen atmosphere, an active climate and meteorological cycles where conditions are such that the working fluid, methane, plays the role that water does on Earth. Titan's surface, with lakes and seas, broad river valleys, sand dunes and mountains was formed by processes like those that have shaped the Earth. Supporting this panoply of Earth-like processes is an ice crust that floats atop what might be a liquid water ocean. Furthermore, Titan is rich in very many different organic compounds—more so than any place in the solar system, except Earth. The Titan Saturn System Mission (TSSM) concept that followed the 2007 TandEM ESA CV proposal [2] and the 2007 Titan Explorer NASA Flagship study [3], was examined [4,5] and prioritized by NASA and ESA in February 2009 as a mission to follow the Europa Jupiter System Mission. The TSSM study, like others before it, again concluded that an orbiter, a montgolfiere hot-air balloon and a surface package (e.g. lake lander, Geosaucer (instrumented heat shield), …) are very high priority elements for any future mission to Titan. Such missions could be conceived as Flagship/Cosmic Vision L-Class or as individual smaller missions that could possibly fit into NASA New Frontiers or ESA Cosmic Vision M-Class budgets. As a result of a multitude of Titan mission studies, a clear blueprint has been laid out for the work needed to reduce the risks inherent in such missions and the areas where advances would be beneficial for elements critical to future Titan missions have been identified. The purpose of this paper is to provide a brief overview of the flagship mission architecture and

  9. Advanced fuel cells for transportation applications. Final report

    SciTech Connect

    1998-02-10

    This Research and Development (R and D) contract was directed at developing an advanced technology compressor/expander for supplying compressed air to Proton Exchange Membrane (PEM) fuel cells in transportation applications. The objective of this project was to develop a low-cost high-efficiency long-life lubrication-free integrated compressor/expander utilizing scroll technology. The goal of this compressor/expander was to be capable of providing compressed air over the flow and pressure ranges required for the operation of 50 kW PEM fuel cells in transportation applications. The desired ranges of flow, pressure, and other performance parameters were outlined in a set of guidelines provided by DOE. The project consisted of the design, fabrication, and test of a prototype compressor/expander module. The scroll CEM development program summarized in this report has been very successful, demonstrating that scroll technology is a leading candidate for automotive fuel cell compressor/expanders. The objectives of the program are: develop an integrated scroll CEM; demonstrate efficiency and capacity goals; demonstrate manufacturability and cost goals; and evaluate operating envelope. In summary, while the scroll CEM program did not demonstrate a level of performance as high as the DOE guidelines in all cases, it did meet the overriding objectives of the program. A fully-integrated, low-cost CEM was developed that demonstrated high efficiency and reliable operation throughout the test program. 26 figs., 13 tabs.

  10. Advanced Coal-Fueled Gas Turbine Program. Final report

    SciTech Connect

    Horner, M.W.; Ekstedt, E.E.; Gal, E.; Jackson, M.R.; Kimura, S.G.; Lavigne, R.G.; Lucas, C.; Rairden, J.R.; Sabla, P.E.; Savelli, J.F.; Slaughter, D.M.; Spiro, C.L.; Staub, F.W.

    1989-02-01

    The objective of the original Request for Proposal was to establish the technological bases necessary for the subsequent commercial development and deployment of advanced coal-fueled gas turbine power systems by the private sector. The offeror was to identify the specific application or applications, toward which his development efforts would be directed; define and substantiate the technical, economic, and environmental criteria for the selected application; and conduct such component design, development, integration, and tests as deemed necessary to fulfill this objective. Specifically, the offeror was to choose a system through which ingenious methods of grouping subcomponents into integrated systems accomplishes the following: (1) Preserve the inherent power density and performance advantages of gas turbine systems. (2) System must be capable of meeting or exceeding existing and expected environmental regulations for the proposed application. (3) System must offer a considerable improvement over coal-fueled systems which are commercial, have been demonstrated, or are being demonstrated. (4) System proposed must be an integrated gas turbine concept, i.e., all fuel conditioning, all expansion gas conditioning, or post-expansion gas cleaning, must be integrated into the gas turbine system.

  11. Advanced coal gasifier-fuel cell power plant systems design

    NASA Technical Reports Server (NTRS)

    Heller, M. E.

    1983-01-01

    Two advanced, high efficiency coal-fired power plants were designed, one utilizing a phosphoric acid fuel cell and one utilizing a molten carbonate fuel cell. Both incorporate a TRW Catalytic Hydrogen Process gasifier and regenerator. Both plants operate without an oxygen plant and without requiring water feed; they, instead, require makeup dolomite. Neither plant requires a shift converter; neither plant has heat exchangers operating above 1250 F. Both plants have attractive efficiencies and costs. While the molten carbonate version has a higher (52%) efficiency than the phosphoric acid version (48%), it also has a higher ($0.078/kWh versus $0.072/kWh) ten-year levelized cost of electricity. The phosphoric acid fuel cell power plant is probably feasible to build in the near term: questions about the TRW process need to be answered experimentally, such as weather it can operate on caking coals, and how effective the catalyzed carbon-dioxide acceptor will be at pilot scale, both in removing carbon dioxide and in removing sulfur from the gasifier.

  12. 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.

  13. 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).

  14. The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

    DOE PAGESBeta

    Ward, Andrew; Downar, Thomas J.; Xu, Y.; March-Leuba, Jose A; Thurston, Carl; Hudson, Nathanael H.; Ireland, A.; Wysocki, A.

    2015-04-22

    The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less

  15. The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

    SciTech Connect

    Ward, Andrew; Downar, Thomas J.; Xu, Y.; March-Leuba, Jose A; Thurston, Carl; Hudson, Nathanael H.; Ireland, A.; Wysocki, A.

    2015-04-22

    The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, the capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.

  16. The Adoption of Advanced Fuel Cycle Technology Under a Single Repository Policy

    SciTech Connect

    Paul Wilson

    2009-11-02

    Develops the tools to investiage the hypothesis that the savings in repository space associated with the implementation of advanced nuclear fuel cycles can result in sufficient cost savings to offset the higher costs of those fuel cycles.

  17. Modeling Constituent Redistribution in U-Pu-Zr Metallic Fuel Using the Advanced Fuel Performance Code BISON

    SciTech Connect

    Douglas Porter; Steve Hayes; Various

    2014-06-01

    The Advanced Fuels Campaign (AFC) metallic fuels currently being tested have higher zirconium and plutonium concentrations than those tested in the past in EBR reactors. Current metal fuel performance codes have limitations and deficiencies in predicting AFC fuel performance, particularly in the modeling of constituent distribution. No fully validated code exists due to sparse data and unknown modeling parameters. Our primary objective is to develop an initial analysis tool by incorporating state-of-the-art knowledge, constitutive models and properties of AFC metal fuels into the MOOSE/BISON (1) framework in order to analyze AFC metallic fuel tests.

  18. Advanced proton-exchange materials for energy efficient fuel cells.

    SciTech Connect

    Fujimoto, Cy H.; Grest, Gary Stephen; Hickner, Michael A.; Cornelius, Christopher James; Staiger, Chad Lynn; Hibbs, Michael R.

    2005-12-01

    The ''Advanced Proton-Exchange Materials for Energy Efficient Fuel Cells'' Laboratory Directed Research and Development (LDRD) project began in October 2002 and ended in September 2005. This LDRD was funded by the Energy Efficiency and Renewable Energy strategic business unit. The purpose of this LDRD was to initiate the fundamental research necessary for the development of a novel proton-exchange membranes (PEM) to overcome the material and performance limitations of the ''state of the art'' Nafion that is used in both hydrogen and methanol fuel cells. An atomistic modeling effort was added to this LDRD in order to establish a frame work between predicted morphology and observed PEM morphology in order to relate it to fuel cell performance. Significant progress was made in the area of PEM material design, development, and demonstration during this LDRD. A fundamental understanding involving the role of the structure of the PEM material as a function of sulfonic acid content, polymer topology, chemical composition, molecular weight, and electrode electrolyte ink development was demonstrated during this LDRD. PEM materials based upon random and block polyimides, polybenzimidazoles, and polyphenylenes were created and evaluated for improvements in proton conductivity, reduced swelling, reduced O{sub 2} and H{sub 2} permeability, and increased thermal stability. Results from this work reveal that the family of polyphenylenes potentially solves several technical challenges associated with obtaining a high temperature PEM membrane. Fuel cell relevant properties such as high proton conductivity (>120 mS/cm), good thermal stability, and mechanical robustness were demonstrated during this LDRD. This report summarizes the technical accomplishments and results of this LDRD.

  19. Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

    NASA Astrophysics Data System (ADS)

    Mella, R.; Wenman, M. R.

    2013-06-01

    Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Advanced Fuel Cycle Economic Tools, Algorithms, and Methodologies

    SciTech Connect

    David E. Shropshire

    2009-05-01

    The Advanced Fuel Cycle Initiative (AFCI) Systems Analysis supports engineering economic analyses and trade-studies, and requires a requisite reference cost basis to support adequate analysis rigor. In this regard, the AFCI program has created a reference set of economic documentation. The documentation consists of the “Advanced Fuel Cycle (AFC) Cost Basis” report (Shropshire, et al. 2007), “AFCI Economic Analysis” report, and the “AFCI Economic Tools, Algorithms, and Methodologies Report.” Together, these documents provide the reference cost basis, cost modeling basis, and methodologies needed to support AFCI economic analysis. The application of the reference cost data in the cost and econometric systems analysis models will be supported by this report. These methodologies include: the energy/environment/economic evaluation of nuclear technology penetration in the energy market—domestic and internationally—and impacts on AFCI facility deployment, uranium resource modeling to inform the front-end fuel cycle costs, facility first-of-a-kind to nth-of-a-kind learning with application to deployment of AFCI facilities, cost tradeoffs to meet nuclear non-proliferation requirements, and international nuclear facility supply/demand analysis. The economic analysis will be performed using two cost models. VISION.ECON will be used to evaluate and compare costs under dynamic conditions, consistent with the cases and analysis performed by the AFCI Systems Analysis team. Generation IV Excel Calculations of Nuclear Systems (G4-ECONS) will provide static (snapshot-in-time) cost analysis and will provide a check on the dynamic results. In future analysis, additional AFCI measures may be developed to show the value of AFCI in closing the fuel cycle. Comparisons can show AFCI in terms of reduced global proliferation (e.g., reduction in enrichment), greater sustainability through preservation of a natural resource (e.g., reduction in uranium ore depletion), value from

  5. 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)

  6. 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.

  7. 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)

  8. Compatibility of alternative fuels with advanced automotive gas turbine and stirling engines. A literature survey

    NASA Technical Reports Server (NTRS)

    Cairelli, J.; Horvath, D.

    1981-01-01

    The application of alternative fuels in advanced automotive gas turbine and Stirling engines is discussed on the basis of a literature survey. These alternative engines are briefly described, and the aspects that will influence fuel selection are identified. Fuel properties and combustion properties are discussed, with consideration given to advanced materials and components. Alternative fuels from petroleum, coal, oil shale, alcohol, and hydrogen are discussed, and some background is given about the origin and production of these fuels. Fuel requirements for automotive gas turbine and Stirling engines are developed, and the need for certain reseach efforts is discussed. Future research efforts planned at Lewis are described.

  9. 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.

  10. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    SciTech Connect

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  11. 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

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

    SciTech Connect

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

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

  13. Enabling Advanced Modeling and Simulations for Fuel-Flexible Combustors

    SciTech Connect

    Pitsch, Heinz

    2010-05-31

    The overall goal of the present project is to enable advanced modeling and simulations for the design and optimization of fuel-flexible turbine combustors. For this purpose we use a high fidelity, extensively-tested large-eddy simulation (LES) code and state-of-the-art models for premixed/partially-premixed turbulent combustion developed in the PI's group. In the frame of the present project, these techniques are applied, assessed, and improved for hydrogen enriched premixed and partially premixed gas-turbine combustion. Our innovative approaches include a completely consistent description of flame propagation; a coupled progress variable/level set method to resolve the detailed flame structure, and incorporation of thermal-diffusion (non-unity Lewis number) effects. In addition, we have developed a general flamelet-type transformation holding in the limits of both non-premixed and premixed burning. As a result, a model for partially premixed combustion has been derived. The coupled progress variable/level method and the general flamelet transformation were validated by LES of a lean-premixed low-swirl burner that has been studied experimentally at Lawrence Berkeley National Laboratory. The model is extended to include the non-unity Lewis number effects, which play a critical role in fuel-flexible combustor with high hydrogen content fuel. More specifically, a two-scalar model for lean hydrogen and hydrogen-enriched combustion is developed and validated against experimental and direct numerical simulation (DNS) data. Results are presented to emphasize the importance of non-unity Lewis number effects in the lean-premixed low-swirl burner of interest in this project. The proposed model gives improved results, which shows that the inclusion of the non-unity Lewis number effects is essential for accurate prediction of the lean-premixed low-swirl flame.

  14. Enabling Advanced Modeling and Simulations for Fuel-Flexible Combustors

    SciTech Connect

    Heinz Pitsch

    2010-05-31

    The overall goal of the present project is to enable advanced modeling and simulations for the design and optimization of fuel-flexible turbine combustors. For this purpose we use a high-fidelity, extensively-tested large-eddy simulation (LES) code and state-of-the-art models for premixed/partially-premixed turbulent combustion developed in the PI's group. In the frame of the present project, these techniques are applied, assessed, and improved for hydrogen enriched premixed and partially premixed gas-turbine combustion. Our innovative approaches include a completely consistent description of flame propagation, a coupled progress variable/level set method to resolve the detailed flame structure, and incorporation of thermal-diffusion (non-unity Lewis number) effects. In addition, we have developed a general flamelet-type transformation holding in the limits of both non-premixed and premixed burning. As a result, a model for partially premixed combustion has been derived. The coupled progress variable/level method and the general flamelet tranformation were validated by LES of a lean-premixed low-swirl burner that has been studied experimentally at Lawrence Berkeley National Laboratory. The model is extended to include the non-unity Lewis number effects, which play a critical role in fuel-flexible combustor with high hydrogen content fuel. More specifically, a two-scalar model for lean hydrogen and hydrogen-enriched combustion is developed and validated against experimental and direct numerical simulation (DNS) data. Results are presented to emphasize the importance of non-unity Lewis number effects in the lean-premixed low-swirl burner of interest in this project. The proposed model gives improved results, which shows that the inclusion of the non-unity Lewis number effects is essential for accurate prediction of the lean-premixed low-swirl flame.

  15. 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

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

    SciTech Connect

    Unal, Cetin; Pasamehmetoglu, Kemal; Carmack, Jon

    2010-01-01

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

  17. 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.

  18. 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

  19. 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.

  20. Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles

    SciTech Connect

    Hardin, Ernest; Blink, James; Carter, Joe; Massimiliano, Fratoni; Greenberg, Harris; Howard, Rob L

    2011-01-01

    The current posture of the used nuclear fuel management program in the U.S. following termination of the Yucca Mountain Project, is to pursue research and development (R&D) of generic (i.e., non-site specific) technologies for storage, transportation and disposal. Disposal R&D is directed toward understanding and demonstrating the performance of reference geologic disposal concepts selected to represent the current state-of-the-art in geologic disposal. One of the principal constraints on waste packaging and emplacement in a geologic repository is management of the waste-generated heat. This paper describes the selection of reference disposal concepts, and thermal management strategies for waste from advanced fuel cycles. A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE) Used Fuel Disposition Campaign, for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. We performed thermal analysis of these concepts using waste inventory cases representing a range of advanced fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress and previous experience in the U.S. repository program. All of the disposal concepts selected for this study use enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. The encapsulating materials (typically clay-based or rock salt) have low intrinsic permeability and plastic rheology that closes voids so that low permeability is maintained. Uniformly low permeability also contributes to chemically reducing conditions common in soft clay, shale, and salt formations. Enclosed modes are associated

  1. 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)

  2. 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.

  3. 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.

  4. 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.

  5. Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011

    SciTech Connect

    Not Listed

    2011-11-01

    The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

  6. 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)

  7. 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.

  8. Alternative Fuels and Advanced Vehicles: Resources for Fleet Managers (Clean Cities) (Presentation)

    SciTech Connect

    Brennan, A.

    2011-04-01

    A discussion of the tools and resources on the Clean Cities, Alternative Fuels and Advanced Vehicles Data Center, and the FuelEconomy.gov Web sites that can help vehicle fleet managers make informed decisions about implementing strategies to reduce gasoline and diesel fuel use.

  9. Geospatial Analysis and Optimization of Fleet Logistics to Exploit Alternative Fuels and Advanced Transportation Technologies: Preprint

    SciTech Connect

    Sparks, W.; Singer, M.

    2010-06-01

    This paper describes how the National Renewable Energy Laboratory (NREL) is developing geographical information system (GIS) tools to evaluate alternative fuel availability in relation to garage locations and to perform automated fleet-wide optimization to determine where to deploy alternative fuel and advanced technology vehicles and fueling infrastructure.

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

    SciTech Connect

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

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

  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. Econometric comparisons of liquid rocket engines for dual-fuel advanced earth-to-orbit shuttles

    NASA Technical Reports Server (NTRS)

    Martin, J. A.

    1978-01-01

    Econometric analyses of advanced Earth-to-orbit vehicles indicate that there are economic benefits from development of new vehicles beyond the space shuttle as traffic increases. Vehicle studies indicate the advantage of the dual-fuel propulsion in single-stage vehicles. This paper shows the economic effect of incorporating dual-fuel propulsion in advanced vehicles. Several dual-fuel propulsion systems are compared to a baseline hydrogen and oxygen system.

  13. Masters Study in Advanced Energy and Fuels Management

    SciTech Connect

    Mondal, Kanchan

    2014-12-08

    There are currently three key drivers for the US energy sector a) increasing energy demand and b) environmental stewardship in energy production for sustainability and c) general public and governmental desire for domestic resources. These drivers are also true for energy nation globally. As a result, this sector is rapidly diversifying to alternate sources that would supplement or replace fossil fuels. These changes have created a need for a highly trained workforce with a the understanding of both conventional and emerging energy resources and technology to lead and facilitate the reinvention of the US energy production, rational deployment of alternate energy technologies based on scientific and business criteria while invigorating the overall economy. In addition, the current trends focus on the the need of Science, Technology, Engineering and Math (STEM) graduate education to move beyond academia and be more responsive to the workforce needs of businesses and the industry. The SIUC PSM in Advanced Energy and Fuels Management (AEFM) program was developed in response to the industries stated need for employees who combine technical competencies and workforce skills similar to all PSM degree programs. The SIUC AEFM program was designed to provide the STEM graduates with advanced technical training in energy resources and technology while simultaneously equipping them with the business management skills required by professional employers in the energy sector. Technical training include core skills in energy resources, technology and management for both conventional and emerging energy technologies. Business skills training include financial, personnel and project management. A capstone internship is also built into the program to train students such that they are acclimatized to the real world scenarios in research laboratories, in energy companies and in government agencies. The current curriculum in the SIUC AEFM will help fill the need for training both recent

  14. Integration of geometric modeling and advanced finite element preprocessing

    NASA Technical Reports Server (NTRS)

    Shephard, Mark S.; Finnigan, Peter M.

    1987-01-01

    The structure to a geometry based finite element preprocessing system is presented. The key features of the system are the use of geometric operators to support all geometric calculations required for analysis model generation, and the use of a hierarchic boundary based data structure for the major data sets within the system. The approach presented can support the finite element modeling procedures used today as well as the fully automated procedures under development.

  15. Performance of the fissionTPC and the Potential to Advance the Thorium Fuel Cycle

    NASA Astrophysics Data System (ADS)

    Towell, Rusty; Niffte Collaboration

    2014-09-01

    The NIFFTE fission Time Projection Chamber (fissionTPC) is a powerful tool that is being developed to take precision measurements of neutron-induced fission cross sections of transuranic elements. During the last run at the Los Alamos Neutron Science Center (LANSCE) the fully instrumented TPC took data for the first time. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow cross section systematics to be controlled at the level of 1%. The fissionTPC performance from this run will be shared. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors including the abundance of thorium verses uranium, minimizing radioactive waste, improved reactor safety, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium fueled nuclear reactors will also be discussed.

  16. Fuel Injector Patternation Evaluation in Advanced Liquid-Fueled, High Pressure, Gas Turbine Combustors, Using Nonintrusive Optical Diagnostic Techniques

    NASA Technical Reports Server (NTRS)

    Locke, R. J.; Hicks, Y. R.; Anderson, R. C.; Zaller, M. M.

    1998-01-01

    Planar laser-induced fluorescence (PLIF) imaging and planar Mie scattering are used to examine the fuel distribution pattern (patternation) for advanced fuel injector concepts in kerosene burning, high pressure gas turbine combustors. Three diverse fuel injector concepts for aerospace applications were investigated under a broad range of operating conditions. Fuel PLIF patternation results are contrasted with those obtained by planar Mie scattering. Further comparison is also made for one injector with data obtained through phase Doppler measurements. Differences in spray patterns for diverse conditions and fuel injector configurations are readily discernible. An examination of the data has shown that a direct determination of the fuel spray angle at realistic conditions is also possible. The results obtained in this study demonstrate the applicability and usefulness of these nonintrusive optical techniques for investigating fuel spray patternation under actual combustor conditions.

  17. Grouped actinide separation in advanced nuclear fuel cycles

    SciTech Connect

    Glatz, J.P.; Malmbeck, R.; Ougier, M.; Soucek, P.; Murakamin, T.; Tsukada, T.; Koyama, T.

    2013-07-01

    Aiming at cleaner waste streams (containing only the short-lived fission products) a partitioning and transmutation (P-T) scheme can significantly reduce the quantities of long-lived radionuclides consigned to waste. Many issues and options are being discussed and studied at present in view of selecting the optimal route. The choice is between individual treatment of the relevant elements and a grouped treatment of all actinides together. In the European Collaborative Project ACSEPT (Actinide recycling by Separation and Transmutation), grouped separation options derived from an aqueous extraction or from a dry pyroprocessing route were extensively investigated. Successful demonstration tests for both systems have been carried out in the frame of this project. The aqueous process called GANEX (Grouped Actinide Extraction) is composed of 2 cycles, a first one to recover the major part of U followed by a co-extraction of Np, Pu, Am, and Cm altogether. The pyro-reprocessing primarily applicable to metallic fuels such as the U-Pu-Zr alloy originally developed by the Argonne National Laboratory (US) in the mid 1980s, has also been applied to the METAPHIX fuels containing up to 5% of minor actinides and 5% of lanthanides (e.g. U{sub 60}Pu{sub 20}-Zr{sub 10}Am{sub 2}Nd{sub 3.5}Y{sub 0.5}Ce{sub 0.5}Gd{sub 0.5}). A grouped actinide separation has been successfully carried out by electrorefining on solid Al cathodes. At present the recovery of the actinides from the alloy formed with Al upon electrodeposition is under investigation, because an efficient P-T cycle requires multiple re-fabrication and re-irradiation. (authors)

  18. 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.

  19. 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.

  20. Advances in atomic data for neutron-capture elements

    NASA Astrophysics Data System (ADS)

    Sterling, Nicholas C.; Witthoeft, Michael C.; Esteves, David A.; Stancil, Phillip C.; Kilcoyne, A. L. David; Bilodeau, Rene C.; Aguilar, Alejandro

    2012-08-01

    Neutron(n)-capture elements (atomic number Z > 30), which can be produced in planetary nebula (PN) progenitor stars via s-process nucleosynthesis, have been detected in nearly 100 PNe. This demonstrates that nebular spectroscopy is a potentially powerful tool for studying the production and chemical evolution of trans-iron elements. However, significant challenges must be addressed before this goal can be achieved. One of the most substantial hurdles is the lack of atomic data for n-capture elements, particularly that needed to solve for their ionization equilibrium (and hence to convert ionic abundances to elemental abundances). To address this need, we have computed photoionization cross sections and radiative and dielectronic recombination rate coefficients for the first six ions of Se and Kr. The calculations were benchmarked against experimental photoionization cross section measurements. In addition, we computed charge transfer (CT) rate coefficients for ions of six n-capture elements. These efforts will enable the accurate determination of nebular Se and Kr abundances, allowing robust investigations of s-process enrichments in PNe.

  1. High-Level Functional and Operational Requirements for the Advanced Fuel Cycle Facilty

    SciTech Connect

    Charles Park

    2006-12-01

    High-Level Functional & Operational Requirements for the AFCF -This document describes the principal functional and operational requirements for the proposed Advanced Fuel Cycle Facility (AFCF). The AFCF is intended to be the world's foremost facility for nuclear fuel cycle research, technology development, and demonstration. The facility will also support the near-term mission to develop and demonstrate technology in support of fuel cycle needs identified by industry, and the long-term mission to retain and retain U.S. leadership in fuel cycle operations. The AFCF is essential to demonstrate a more proliferation-resistant fuel cycle and make long-term improvements in fuel cycle effectiveness, performance and economy.

  2. Advanced turbine design for coal-fueled engines

    NASA Astrophysics Data System (ADS)

    Bornstein, N. S.

    1992-07-01

    The objective of this task is to perform a technical assessment of turbine blading for advanced second generation PFBC conditions, identify specific problems/issues, and recommend an approach for solving any problems identified. A literature search was conducted, problems associated with hot corrosion defined and limited experiments performed. Sulfidation corrosion occurs in industrial, marine and aircraft gas turbine engines and is due to the presence of condensed alkali (sodium) sulfates. The principle source of the alkali in industrial, marine and aircraft gas turbine engines is sea salt crystals. The principle source of the sulfur is not the liquid fuels, but the same ocean born crystals. Moreover deposition of the corrosive salt occurs primarily by a non-equilibrium process. Sodium will be present in the cleaned combusted gases that enter the PFBC turbine. Although equilibrium condensation is not favored, deposition via impaction is probable. Marine gas turbines operate in sodium chloride rich environments without experiencing the accelerated attack noted in coal fired boilers where condensed chlorides contact metallic surfaces. The sulfates of calcium and magnesium are the products of the reactions used to control sulfur. Based upon industrial gas turbine experience and laboratory tests, calcium and magnesium sulfates are, at temperatures up to 1500 F (815 C), relatively innocuous salts. In this study it is found that at 1650 F (900 C) and above, calcium sulfate becomes an aggressive corrodent.

  3. Development of advanced fuel cell system, phase 3

    NASA Technical Reports Server (NTRS)

    Handley, L. M.; Meyer, A. P.; Bell, W. F.

    1975-01-01

    A multiple task research and development program was performed to improve the weight, life, and performance characteristics of hydrogen-oxygen alkaline fuel cells for advanced power systems. Gradual wetting of the anode structure and subsequent long-term performance loss was determined to be caused by deposition of a silicon-containing material on the anode. This deposit was attributed to degradation of the asbestos matrix, and attention was therefore placed on development of a substitute matrix of potassium titanate. An 80 percent gold 20 percent platinum catalyst cathode was developed which has the same performance and stability as the standard 90 percent gold - 10 percent platinum cathode but at half the loading. A hybrid polysulfone/epoxy-glass fiber frame was developed which combines the resistance to the cell environment of pure polysulfone with the fabricating ease of epoxy-glass fiber laminate. These cell components were evaluated in various configurations of full-size cells. The ways in which the baseline engineering model system would be modified to accommodate the requirements of the space tug application are identified.

  4. 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.

  5. Recent advances in hybrid/mixed finite elements

    NASA Technical Reports Server (NTRS)

    Pian, T. H. H.

    1985-01-01

    In formulations of Hybrid/Mixed finite element methods respectively by the Hellinger-Reissner principle and the Hu-Washizu principle, the stress equilibrium equations are brought in as conditions of constraint through the introduction of additional internal displacement parameters. These two approaches are more flexible and have better computing efficiencies. A procedure for the choice of assumed stress terms for 3-D solids is suggested. Example solutions are given for plates and shells using the present formulations and the idea of semiloof elements.

  6. Closing the US Fuel Cycle: Siting Considerations for the Global Nuclear Energy Partnership Facilities - Siting the Advanced Fuel Cycle Facility

    SciTech Connect

    Griffith, A.; Boger, J.; Perry, J.

    2008-07-01

    The Global Nuclear Energy Partnership (GNEP), launched in February, 2006, proposes to introduce used nuclear fuel recycling in the United States (U.S.) with improved proliferation-resistance and a more effective waste management approach. This program is evaluating ways to close the fuel cycle in a manner that introduces the most advanced technologies of today and builds on recent breakthroughs in U.S. national laboratories while drawing on international and industry partnerships. Central to moving this advanced fuel recycling technology from the laboratory to commercial implementation is the development and siting of three proposed GNEP facilities: the Consolidated Fuel Treatment Center (CFTC), the Advanced Burner Reactor (ABR), and the Advanced Fuel Cycle Facility (AFCF). These three projects are envisioned to introduce used fuel separations, advanced fuel fabrication, and fast reactor technology in a manner that efficiently recycles material, produces the most energy out of the existing inventory of used fuel, and improves our ability to manage nuclear waste. The CFTC and ABR are sited under GNEP but will depend on industry involvement and will not be covered by this paper. This paper will cover considerations for siting the AFCF. The AFCF will provide the U.S. with the capabilities required to evaluate technologies that separate used fuel into reusable material and waste in a proliferation-resistant manner. The separations technology demonstration capability is coupled with a remote transmutation fuel fabrication demonstration capability in an integrated manner that demonstrates advanced safeguard technologies. In conclusion: As a flexible, multi-purpose demonstration facility, the AFCF will provide the U.S. with a powerful and unique capability to quickly bring innovative nuclear fuel recycling technology from the laboratory to the commercial market with high confidence. The siting of AFCF capabilities at one or more of the six DOE laboratories being evaluated

  7. 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.

  8. 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.

  9. Concept of advanced spent fuel reprocessing based on ion exchange

    SciTech Connect

    Suzuki, Tatsuya; Takahashi, Kazuyuki; Nogami, Masanobu; Nomura, Masao; Fujii, Yasuhiko; Ozawa, Masaki |; Koyama, Shinichi; Mimura, Hitosi; Fujita, Reiko

    2007-07-01

    . Furthermore, the ion exchange is appropriate for multi-element mutual separation rather than single element extraction. In the future, ion exchange reprocessing would be expected to be the comprehensive separation process for spent fuels to recover precious and usable elements and to reduce the amount of wastes. (authors)

  10. 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).

  11. 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.

  12. Implementing Common Data Elements Across Studies to Advance Research

    PubMed Central

    Cohen, Marlene Z.; Thompson, Cheryl Bagley; Yates, Bernice; Zimmerman, Lani; Pullen, Carol H.

    2014-01-01

    Challenges arise in building the knowledge needed for evidence based practice partially because obtaining clinical research data is expensive and complicated, and many studies have small sample sizes. Combining data from several studies may have the advantage of increasing the impact of the findings, or expanding the population to which findings may be generalized. The use of common data elements will allow this combining and, in turn, create big data, which is an important approach that may accelerate knowledge development. This article discusses the philosophy of using common data elements across research studies and illustrates their use by the processes in a Developmental Center grant funded by the National Institutes of Health. The researchers identified a set of data elements and used them across several pilot studies. Issues that need to be considered in the adoption and implementation of common data elements across pilot studies include theoretical framework, purpose of the common measures, respondent burden, team work, managing large data sets, grant writing, and unintended consequences. We describe these challenges and solutions that can be implemented to manage them. PMID:25771192

  13. 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.

  14. Renovation of Chemical Processing Facility for Development of Advanced Fast Reactor Fuel Cycle System in JNC

    SciTech Connect

    Atsushi Aoshima; Shigehiko Miyachi; Takashi Suganuma; Shinichi Nemoto

    2002-07-01

    The CPF had 4 laboratories (operation room A, laboratory A, laboratory C and analysis laboratory) in connection with reprocessing technology. The main laboratory, operation room A, has 5 hot cells. Since equipment in the main cell had been designed for small-scale verification of existing reprocessing steps, it was hardly able to respond flexibly to experimental studies on advanced technology. It was decided to remodel the cell according to the design that was newly laid out in order to ensure the function and space to conduct various basic tests. The other laboratories had no glove boxes for conducting basic experiments of important elements in the advanced reprocessing, such as actinides except U and Pu, lanthanides and so on. In order to meet various requirements of innovative technologies on advanced fuel cycle development, one laboratory is established more for study on dry reprocessing, and glove boxes, hoods and analytical equipment such as NMR, FT-IR, TI-MS are newly installed in the other laboratories in this renovation. After the renovation, hot tests in the CPF will be resumed from April 2002. (authors)

  15. Part A - Advanced turbine systems. Part B - Materials/manufacturing element of the Advanced Turbine Systems Program

    SciTech Connect

    Karnitz, M.A.

    1996-06-01

    The DOE Offices of Fossil Energy and Energy Efficiency and Renewable Energy have initiated a program to develop advanced turbine systems for power generation. The objective of the Advanced Turbine Systems (ATS) Program is to develop ultra-high efficiency, environmentally superior, and cost competitive gas turbine systems for utility and industrial applications. One of the supporting elements of the ATS Program is the Materials/Manufacturing Technologies Task. The objective of this element is to address the critical materials and manufacturing issues for both industrial and utility gas turbines.

  16. Fuel economy screening study of advanced automotive gas turbine engines

    NASA Technical Reports Server (NTRS)

    Klann, J. L.

    1980-01-01

    Fuel economy potentials were calculated and compared among ten turbomachinery configurations. All gas turbine engines were evaluated with a continuously variable transmission in a 1978 compact car. A reference fuel economy was calculated for the car with its conventional spark ignition piston engine and three speed automatic transmission. Two promising engine/transmission combinations, using gasoline, had 55 to 60 percent gains over the reference fuel economy. Fuel economy sensitivities to engine design parameter changes were also calculated for these two combinations.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Alternative Fuel and Advanced Technology Commercial Lawn Equipment (Spanish version); Clean Cities, Energy Efficiency & Renewable Energy (EERE)

    SciTech Connect

    Nelson, Erik

    2015-06-01

    Powering commercial lawn equipment with alternative fuels or advanced engine technology is an effective way to reduce U.S. dependence on petroleum, reduce harmful emissions, and lessen the environmental impacts of commercial lawn mowing. Numerous alternative fuel and fuel-efficient advanced technology mowers are available. Owners turn to these mowers because they may save on fuel and maintenance costs, extend mower life, reduce fuel spillage and fuel theft, and demonstrate their commitment to sustainability.

  3. Design and fabrication of advanced EUV diffractive elements

    SciTech Connect

    Naulleau, Patrick P.; Liddle, J. Alexander; Salmassi, Farhad; Anderson, Erik H.; Gullikson, Eric M.

    2003-11-16

    As extreme ultraviolet (EUV) lithography approaches commercial reality, the development of EUV-compatible diffractive structures becomes increasingly important. Such devices are relevant to many aspects of EUV technology including interferometry, illumination, and spectral filtering. Moreover, the current scarcity of high power EUV sources makes the optical efficiency of these diffractive structures a paramount concern. This fact has led to a strong interest in phase-enhanced diffractive structures. Here we describe recent advancements made in the fabrication of such devices.

  4. Recent anode advances in solid oxide fuel cells

    NASA Astrophysics Data System (ADS)

    Sun, Chunwen; Stimming, Ulrich

    Solid oxide fuel cells (SOFCs) are electrochemical reactors that can directly convert the chemical energy of a fuel gas into electrical energy with high efficiency and in an environment-friendly way. The recent trends in the research of solid oxide fuel cells concern the use of available hydrocarbon fuels, such as natural gas. The most commonly used anode material Ni/YSZ cermet exhibits some disadvantages when hydrocarbons were used as fuels. Thus it is necessary to develop alternative anode materials which display mixed conductivity under fuel conditions. This article reviews the recent developments of anode in SOFCs with principal emphasis on the material aspects. In addition, the mechanism and kinetics of fuel oxidation reactions are also addressed. Various processes used for the cost-effective fabrication of anode have also been summarized. Finally, this review will be concluded with personal perspectives on the future research directions of this area.

  5. 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.

  6. 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.

  7. 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.

  8. 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/.

  9. 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.

  10. Fabrication of synthetic diffractive elements using advanced matrix laser lithography

    NASA Astrophysics Data System (ADS)

    Škereň, M.; Svoboda, J.; Květoň, M.; Fiala, P.

    2013-02-01

    In this paper we present a matrix laser writing device based on a demagnified projection of a micro-structure from a computer driven spatial light modulator. The device is capable of writing completely aperiodic micro-structures with resolution higher than 200 000 DPI. An optical system is combined with ultra high precision piezoelectric stages with an elementary step ~ 4 nm. The device operates in a normal environment, which significantly decreases the costs compared to competitive technologies. Simultaneously, large areas can be exposed up to 100 cm2. The capabilities of the constructed device will be demonstrated on particular elements fabricated for real applications. The optical document security is the first interesting field, where the synthetic image holograms are often combined with sophisticated aperiodic micro-structures. The proposed technology can easily write simple micro-gratings creating the color and kinetic visual effects, but also the diffractive cryptograms, waveguide couplers, and other structures recently used in the field of optical security. A general beam shaping elements and special photonic micro-structures are another important applications which will be discussed in this paper.

  11. Advances in passive imaging elements with micromirror array

    NASA Astrophysics Data System (ADS)

    Maekawa, Satoshi; Nitta, Kouichi; Matoba, Osamu

    2008-02-01

    We have proposed a new passive imaging optics which consists of a grid array of micro roof mirrors working as dihedral corner reflectors. Although this element forms mirror-like images at opposite side of objects, the images are real. Because the imaging principle of the proposed element is based on accumulation of rays, the design of each light path makes many kinds of devices possible. So, we propose two variations of such a device. One device consists of an array of micro retroreflectors and a half mirror, and it can also form real mirror-like images. The advantage of this device is wide range of view, because the displacement of each retororeflector is not limited on a plane unlike the roof mirror grid array. The other consists of an array of long dihedral corner reflectors. Although this structure has been already known as a roof mirror array, it can be used for imaging. This device forms two heterogeneous images. One is real at the same side of an object, and the other is virtual at the opposite side. This is a conjugate imaging optics of a slit mirror array whose mirror surface is perpendicular to the device surface. The advantage of a roor mirror array is that the real image has horizontal parallax and can be seen in air naturally.

  12. Development of Technologies for the Simultaneous Separation of Cesium and Strontium from Spent Nuclear Fuel as Part of an Advanced Fuel Cycle

    SciTech Connect

    Jack D. Law; R. Scott HErbst; David H. Meikrantz; Dean R. Peterman; Catherine L. Riddle; Richard D. Tillotson; Terry A. Todd

    2005-04-01

    As part of the Advanced Fuel Cycle Initiative, two solvent extraction technologies are being developed to simultaneously separate cesium and strontium from dissolved spent nuclear fuel. The first process utilizes a solvent consisting of chlorinated cobalt dicarbollide and polyethylene glycol extractants in a phenyltrifluoromethyl sulfone diluent. Recent improvements to the process include development of a new, non-nitroaromatic diluent and development of new stripping reagents, including a regenerable strip reagent that can be recovered and recycled. Countercurrent flowsheets have been designed and tested on simulated and actual spent nuclear fuel feed streams with both cesium and strontium removal efficiencies of greater than 99 %. The second process developed to simultaneously separate cesium and strontium from spent nuclear fuel is based on two highly-specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. A solvent composition has been developed that enables both elements to be removed together and, in fact, a synergistic effect was observed with strontium distributions in the combined solvent that are much higher that in the strontium extraction (SREX) process. Initial laboratory test results of the new combined cesium and strontium extraction process indicate good extraction and stripping performance. A flowsheet for treatment of spent nuclear fuel is currently being developed.

  13. 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.

  14. 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.

  15. Development of Advanced Hydrocarbon Fuels at Marshall Space Flight Center

    NASA Technical Reports Server (NTRS)

    Bai, S. D.; Dumbacher, P.; Cole, J. W.

    2002-01-01

    This was a small-scale, hot-fire test series to make initial measurements of performance differences of five new liquid fuels relative to rocket propellant-1 (RP-1). The program was part of a high-energy-density materials development at Marshall Space Flight Center (MSFC), and the fuels tested were quadricyclane, 1-7 octodiyne, AFRL-1, biclopropylidene, and competitive impulse noncarcinogenic hypergol (CINCH) (di-methyl-aminoethyl-azide). All tests were conducted at MSFC. The first four fuels were provided by the U.S. Air Force Research Laboratory (AFRL), Edwards Air Force Base, CA. The U.S. Army, Redstone Arsenal, Huntsville, AL, provided the CINCH. The data recorded in all hot-fire tests were used to calculate specific impulse and characteristic exhaust velocity for each fuel, then compared to RP-1 at the same conditions. This was not an exhaustive study, comparing each fuel to RP-1 at an array of mixture ratios, nor did it include important fuel parameters, such as fuel handling or long-term storage. The test hardware was designed for liquid oxygen (lox)/RP-1, then modified for gaseous oxygen/RP-1 to avoid two-phase lox at very small flow rates. All fuels were tested using the same thruster/injector combination designed for RP-1. The results of this test will be used to determine which fuels will be tested in future test programs.

  16. Materials/manufacturing element of the Advanced Turbine Systems Program

    SciTech Connect

    Karnitz, M.A.; Holcomb, R.S.; Wright, I.G.

    1995-10-01

    The technology based portion of the Advanced Turbine Systems Program (ATS) contains several subelements which address generic technology issues for land-based gas-turbine systems. One subelement is the Materials/Manufacturing Technology Program which is coordinated by DOE-Oak Ridge Operations and Oak Ridge National Laboratory (ORNL). The work in this subelement is being performed predominantly by industry with assistance from universities and the national laboratories. Projects in this subelement are aimed toward hastening the incorporation of new materials and components in gas turbines. A materials/manufacturing plan was developed in FY 1994 with input from gas turbine manufacturers, materials suppliers, universities, and government laboratories. The plan outlines seven major subelements which focus on materials issues and manufacturing processes. Work is currently under way in four of the seven major subelements. There are now major projects on coatings and process development, scale-up of single crystal airfoil manufacturing technology, materials characterization, and technology information exchange.

  17. Advanced Theoretical Studies for Chemical Identification of the Heaviest Elements

    NASA Astrophysics Data System (ADS)

    Pershina, V.

    2015-06-01

    Both relativistic Dirac-Fock (DF) correlated, CCSD(T), and the density-functional theory (DFT) methods with various GGA exchange-correlation potentials agree on an increase in the M-Au bond strength from M = Cn to Fl. This means that the adsorption energy of Fl on gold, as measured by gas-phase chromatography experiments, must be larger for Fl than for Cn. For the weak (van der Waals) interactions, the trend to an increase in bonding from Cn2 to Fl2 obtained at the DFT level of theory is also confirmed by the DF CCSD(T) calculations. Thus, for comparative studies, relativistic DFT is a very reliable and efficient tool. First calculations of adsorption of TlOH and element 113OH on a hydroxylated quartz surface are reported.

  18. 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.

  19. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    SciTech Connect

    Navarro, Jorge

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  20. Advanced Anodes for High-Temperature Fuel Cells

    SciTech Connect

    Atkinson, Alan; Barnett, Scott A.; Gorte, Raymond J.; Irvine, John T.; McEvoy, Augustin J.; Mogensen, Mogens; Singhal, Subhash C.; Vohs, John M.

    2004-01-04

    Fuel cells will undoubtedly find widespread use in this new millenium in the conversion of chemical to electrical energy, as they offer very high efficiencies and have unique scalability in electricity-generation applications. The solid-oxide fuel cell (SOFC) is one of the most exciting of these energy technologies; it is an all-ceramic device that operates at temperatures in the range 500-1000 C. The SOFC offers certain advantages over lower temperature fuel cells, notably its ability to use carbon monoxide as a fuel rather than being poisoned by it, and the availability of high-grade exhaust heat for combined heat and power, or combined cycle gas-turbine applications. Although cost is clearly the most important barrier to widespread SOFC implementation, perhaps the most important technical barriers currently being addressed relate to the electrodes, particularly the fuel electrode or anode. In terms of mitigating global warming, the ability of the SOFC to use commonly available fuels at high efficiency promises an effective and early reduction in carbon dioxide emissions and,hence, is one of the lead new technologies for improving the environment. Here, we discuss recent developments of SOFC fuel electrodes that will enable the better use of readily available fuels.

  1. Advanced anodes for high-temperature fuel cells.

    PubMed

    Atkinson, A; Barnett, S; Gorte, R J; Irvine, J T S; McEvoy, A J; Mogensen, M; Singhal, S C; Vohs, J

    2004-01-01

    Fuel cells will undoubtedly find widespread use in this new millennium in the conversion of chemical to electrical energy, as they offer very high efficiencies and have unique scalability in electricity-generation applications. The solid-oxide fuel cell (SOFC) is one of the most exciting of these energy technologies; it is an all-ceramic device that operates at temperatures in the range 500-1,000 degrees C. The SOFC offers certain advantages over lower temperature fuel cells, notably its ability to use carbon monoxide as a fuel rather than being poisoned by it, and the availability of high-grade exhaust heat for combined heat and power, or combined cycle gas-turbine applications. Although cost is clearly the most important barrier to widespread SOFC implementation, perhaps the most important technical barriers currently being addressed relate to the electrodes, particularly the fuel electrode or anode. In terms of mitigating global warming, the ability of the SOFC to use commonly available fuels at high efficiency, promises an effective and early reduction in carbon dioxide emissions, and hence is one of the lead new technologies for improving the environment. Here, we discuss recent developments of SOFC fuel electrodes that will enable the better use of readily available fuels. PMID:14704781

  2. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2006-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  3. 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.

  4. 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.

  5. Advanced Petroleum-Based Fuels -- Diesel Emissions Control Project (APBF-DEC)

    SciTech Connect

    Not Available

    2003-03-01

    Annual progress report of the Advanced Petroleum-based fuels-Diesel Emissions Control Project. Contains information on 5 test projects to determine the best combinations of low-sulfur diesel fuels, lubricants, diesel engines, and emission control systems to meet projected emissions standards.

  6. Advances in Materials and System Technology for Portable Fuel Cells

    NASA Technical Reports Server (NTRS)

    Narayanan, Sekharipuram R.

    2007-01-01

    This viewgraph presentation describes the materials and systems engineering used for portable fuel cells. The contents include: 1) Portable Power; 2) Technology Solution; 3) Portable Hydrogen Systems; 4) Direct Methanol Fuel Cell; 5) Direct Methanol Fuel Cell System Concept; 6) Overview of DMFC R&D at JPL; 7) 300-Watt Portable Fuel Cell for Army Applications; 8) DMFC units from Smart Fuel Cell Inc, Germany; 9) DMFC Status and Prospects; 10) Challenges; 11) Rapid Screening of Well-Controlled Catalyst Compositions; 12) Screening of Ni-Zr-Pt-Ru alloys; 13) Issues with New Membranes; 14) Membranes With Reduced Methanol Crossover; 15) Stacks; 16) Hybrid DMFC System; 17) Small Compact Systems; 18) Durability; and 19) Stack and System Parameters for Various Applications.

  7. Applications study of advanced power generation systems utilizing coal-derived fuels, volume 2

    NASA Technical Reports Server (NTRS)

    Robson, F. L.

    1981-01-01

    Technology readiness and development trends are discussed for three advanced power generation systems: combined cycle gas turbine, fuel cells, and magnetohydrodynamics. Power plants using these technologies are described and their performance either utilizing a medium-Btu coal derived fuel supplied by pipeline from a large central coal gasification facility or integrated with a gasification facility for supplying medium-Btu fuel gas is assessed.

  8. 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.

  9. 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.

  10. 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.

  11. Application of advanced material systems to composite frame elements

    NASA Technical Reports Server (NTRS)

    Llorente, Steven; Minguet, Pierre; Fay, Russell; Medwin, Steven

    1992-01-01

    A three phase program has been conducted to investigate DuPont's Long Discontinuous Fiber (LDF) composites. Additional tests were conducted to compare LDF composites against toughened thermosets and a baseline thermoset system. Results have shown that the LDF AS4/PEKK offers improved interlaminar (flange bending) strength with little reduction in mechanical properties due to the discontinuous nature of the fibers. In the third phase, a series of AS4/PEKK LDF C-section curved frames (representing a typical rotorcraft light frame) were designed, manufactured and tested. Specimen reconsolidation after 'stretch forming' and frame thickness were found to be key factors in this light frame's performance. A finite element model was constructed to correlate frame test results with expected strain levels determined from material property tests. Adequately reconsolidated frames performed well and failed at strain levels at or above baseline thermoset material test strains. Finally a cost study was conducted which has shown that the use of LDF for this frame would result in a significant cost savings, for moderate to large lot sizes compared with the hand lay-up of a thermoset frame.

  12. Numerical design of advanced multi-element airfoils

    NASA Technical Reports Server (NTRS)

    Mathias, Donovan L.; Cummings, Russell M.

    1994-01-01

    The current study extends the application of computational fluid dynamics to three-dimensional high-lift systems. Structured, overset grids are used in conjunction with an incompressible Navier-Stokes flow solver to investigate flow over a two-element high-lift configuration. The computations were run in a fully turbulent mode using the one-equation Baldwin-Barth turbulence model. The geometry consisted of an unswept wing which spanned a wind tunnel test section. Flows over full and half-span Fowler flap configurations were computed. Grid resolution issues were investigated in two dimensional studies of the flapped airfoil. Results of the full-span flap wing agreed well with experimental data and verified the method. Flow over the wing with the half-span was computed to investigate the details of the flow at the free edge of the flap. The results illustrated changes in flow streamlines, separation locations, and surface pressures due to the vortex shed from the flap edge.

  13. Advances and recent trends in heterogeneous photo(electro)-catalysis for solar fuels and chemicals.

    PubMed

    Highfield, James

    2015-01-01

    In the context of a future renewable energy system based on hydrogen storage as energy-dense liquid alcohols co-synthesized from recycled CO2, this article reviews advances in photocatalysis and photoelectrocatalysis that exploit solar (photonic) primary energy in relevant endergonic processes, viz., H2 generation by water splitting, bio-oxygenate photoreforming, and artificial photosynthesis (CO2 reduction). Attainment of the efficiency (>10%) mandated for viable techno-economics (USD 2.00-4.00 per kg H2) and implementation on a global scale hinges on the development of photo(electro)catalysts and co-catalysts composed of earth-abundant elements offering visible-light-driven charge separation and surface redox chemistry in high quantum yield, while retaining the chemical and photo-stability typical of titanium dioxide, a ubiquitous oxide semiconductor and performance "benchmark". The dye-sensitized TiO2 solar cell and multi-junction Si are key "voltage-biasing" components in hybrid photovoltaic/photoelectrochemical (PV/PEC) devices that currently lead the field in performance. Prospects and limitations of visible-absorbing particulates, e.g., nanotextured crystalline α-Fe2O3, g-C3N4, and TiO2 sensitized by C/N-based dopants, multilayer composites, and plasmonic metals, are also considered. An interesting trend in water splitting is towards hydrogen peroxide as a solar fuel and value-added green reagent. Fundamental and technical hurdles impeding the advance towards pre-commercial solar fuels demonstration units are considered. PMID:25884553

  14. 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.

  15. Advancing Symptom Science Through Use of Common Data Elements

    PubMed Central

    Redeker, Nancy S.; Anderson, Ruth; Bakken, Suzanne; Corwin, Elizabeth; Docherty, Sharron; Dorsey, Susan G.; Heitkemper, Margaret; McCloskey, Donna Jo; Moore, Shirley; Pullen, Carol; Rapkin, Bruce; Schiffman, Rachel; Waldrop-Valverde, Drenna; Grady, Patricia

    2015-01-01

    Background Use of common data elements (CDEs), conceptually defined as variables that are operationalized and measured in identical ways across studies, enables comparison of data across studies in ways that would otherwise be impossible. Although healthcare researchers are increasingly using CDEs, there has been little systematic use of CDEs for symptom science. CDEs are especially important in symptom science because people experience common symptoms across a broad range of health and developmental states, and symptom management interventions may have common outcomes across populations. Purposes The purposes of this article are to (a) recommend best practices for the use of CDEs for symptom science within and across centers; (b) evaluate the benefits and challenges associated with the use of CDEs for symptom science; (c) propose CDEs to be used in symptom science to serve as the basis for this emerging science; and (d) suggest implications and recommendations for future research and dissemination of CDEs for symptom science. Design The National Institute of Nursing Research (NINR)-supported P20 and P30 Center directors applied published best practices, expert advice, and the literature to identify CDEs to be used across the centers to measure pain, sleep, fatigue, and affective and cognitive symptoms. Findings We generated a minimum set of CDEs to measure symptoms. Conclusions The CDEs identified through this process will be used across the NINR Centers and will facilitate comparison of symptoms across studies. We expect that additional symptom CDEs will be added and the list will be refined in future work. Clinical Relevance Journal of Nursing Scholarship, 47:5, ©2015 Sigma Theta Tau International. PMID:26250061

  16. 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)

  17. The BWR advanced fuel design experience using Studsvik CMS

    SciTech Connect

    DiGiovine, A.S.; Gibbon, S.H.; Wiksell, G.

    1996-12-31

    The current trend within the nuclear industry is to maximize generation by extending cycle lengths and taking outages as infrequently as possible. As a result, many utilities have begun to use fuel designed to meet these more demanding requirements. These fuel designs are significantly more heterogeneous in mechanical and neutronic detail than prior designs. The question arises as to how existing in-core fuel management codes, such as Studsvik CMS perform in modeling cores containing these designs. While this issue pertains to both pressurized water reactors (PWRs) and boiling water reactors (BWRs), this summary focuses on BWR applications.

  18. Advanced supersonic technology concept study: Hydrogen fueled configuration

    NASA Technical Reports Server (NTRS)

    Brewer, G. D.

    1974-01-01

    Conceptual designs of hydrogen fueled supersonic transport configurations for the 1990 time period were developed and compared with equivalent technology Jet A-1 fueled vehicles to determine the economic and performance potential of liquid hydrogen as an alternate fuel. Parametric evaluations of supersonic cruise vehicles with varying design and transport mission characteristics established the basis for selecting a preferred configuration which was then studied in greater detail. An assessment was made of the general viability of the selected concept including an evaluation of costs and environmental considerations, i.e., exhaust emissions and sonic boom characteristics. Technology development requirements and suggested implementation schedules are presented.

  19. 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

  20. Advanced Interactive Facades - Critical Elements for Future GreenBuildings?

    SciTech Connect

    Selkowitz, Stephen; Aschehoug, Oyvind; Lee, Eleanor S.

    2003-11-01

    Building designers and owners have always been fascinated with the extensive use of glass in building envelopes. Today the highly glazed facade has almost become an iconic element for a 'green building' that provides daylighting and a visual connection with the natural environment. Even before the current interest in green buildings there was no shortage of highly glazed building designs. But many of these buildings either rejected sunlight, and some associated daylight and view with highly reflective glazings or used highly transmissive glass and encountered serious internal comfort problems that could only be overcome with large HVAC systems, resulting in significant energy, cost and environmental penalties. From the 1960's to the 1990's innovation in glazing made heat absorbing glass, reflective glass and double glazing commonplace, with an associated set of aesthetic features. In the last decade there has been a subtle shift from trying to optimize an ideal, static design solution using these glazings to making the facade responsive, interactive and even intelligent. More sophisticated design approaches and technologies have emerged using new high-performance glazing, improved shading and solar control systems, greater use of automated controls, and integration with other building systems. One relatively new architectural development is the double glass facade that offers a cavity that can provide improved acoustics, better solar control and enhanced ventilation. Taken to its ultimate development, an interactive facade should respond intelligently and reliably to the changing outdoor conditions and internal performance needs. It should exploit available natural energies for lighting, heating and ventilation, should be able to provide large energy savings compared to conventional technologies, and at the same time maintain optimal indoor visual and thermal comfort conditions. As photovoltaic costs decrease in the future, these onsite power systems will be

  1. Fuel Cells for Portable Power: 1. Introduction to DMFCs; 2. Advanced Materials and Concepts for Portable Power Fuel Cells

    SciTech Connect

    Zelenay, Piotr

    2012-07-16

    Thanks to generally less stringent cost constraints, portable power fuel cells, the direct methanol fuel cell (DMFC) in particular, promise earlier market penetration than higher power polymer electrolyte fuel cells (PEFCs) for the automotive and stationary applications. However, a large-scale commercialization of DMFC-based power systems beyond niche applications already targeted by developers will depend on improvements to fuel cell performance and performance durability as well as on the reduction in cost, especially of the portable systems on the higher end of the power spectrum (100-250 W). In this part of the webinar, we will focus on the development of advanced materials (catalysts, membranes, electrode structures, and membrane electrode assemblies) and fuel cell operating concepts capable of fulfilling two key targets for portable power systems: the system cost of $5/W and overall fuel conversion efficiency of 2.0-2.5 kWh/L. Presented research will concentrate on the development of new methanol oxidation catalysts, hydrocarbon membranes with reduced methanol crossover, and improvements to component durability. Time permitted, we will also present a few highlights from the development of electrocatalysts for the oxidation of two alternative fuels for the direct-feed fuel cells: ethanol and dimethyl ether.

  2. Screening of advanced cladding materials and UN-U3Si5 fuel

    NASA Astrophysics Data System (ADS)

    Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

    2015-07-01

    In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in

  3. 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.

  4. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2007-09-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  5. Advanced and In Situ Analytical Methods for Solar Fuel Materials.

    PubMed

    Chan, Candace K; Tüysüz, Harun; Braun, Artur; Ranjan, Chinmoy; La Mantia, Fabio; Miller, Benjamin K; Zhang, Liuxian; Crozier, Peter A; Haber, Joel A; Gregoire, John M; Park, Hyun S; Batchellor, Adam S; Trotochaud, Lena; Boettcher, Shannon W

    2016-01-01

    In situ and operando techniques can play important roles in the development of better performing photoelectrodes, photocatalysts, and electrocatalysts by helping to elucidate crucial intermediates and mechanistic steps. The development of high throughput screening methods has also accelerated the evaluation of relevant photoelectrochemical and electrochemical properties for new solar fuel materials. In this chapter, several in situ and high throughput characterization tools are discussed in detail along with their impact on our understanding of solar fuel materials. PMID:26267386

  6. Advanced coal-fueled industrial cogeneration gas turbine system

    SciTech Connect

    LeCren, R.T.; Cowell, L.H.; Galica, M.A.; Stephenson, M.D.; Wen, C.S.

    1990-07-01

    The objective of the Solar/METC program is to prove the technical, economic, and environmental feasibility of coal-fired gas turbine for cogeneration applications through tests of a Centaur Type H engine system operated on coal fuel throughout the engine design operating range. This quarter, work was centered on design, fabrication, and testing of the combustor, cleanup, fuel specifications, and hot end simulation rig. 2 refs., 59 figs., 29 tabs.

  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. 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.

  10. 2-D Time-Dependent Fuel Element, Thermal Analysis Code System.

    Energy Science and Technology Software Center (ESTSC)

    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

  11. 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.

  12. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    PubMed

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

  13. SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: First Results Report

    SciTech Connect

    Eudy, L.; Chandler, K.

    2011-03-01

    This report describes operations at SunLine Transit Agency for their newest prototype fuel cell bus and five compressed natural gas (CNG) buses. In May 2010, SunLine began operating its sixth-generation hydrogen fueled bus, an Advanced Technology (AT) fuel cell bus that incorporates the latest design improvements to reduce weight and increase reliability and performance. The agency is collaborating with the U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) to evaluate the bus in revenue service. This report provides the early data results and implementation experience of the AT fuel cell bus since it was placed in service.

  14. 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.

  15. Advanced Vehicle Testing Activity: Hydrogen-Fueled Mercedes Sprinter Van Operating Summary - January 2003

    SciTech Connect

    Karner, D.; Francfort, J.E.

    2003-01-22

    Over the past two years, Arizona Public Service, a subsidiary of Pinnacle West Capital Corporation, in cooperation with the U.S. Department of Energy's Advanced Vehicle Testing Activity, tested four gaseous fuel vehicles as part of its alternative fueled vehicle fleet. One vehicle operated initially using compressed natural gas (CNG) and later a blend of CNG and hydrogen. Of the other three vehicles, one was fueled with pure hydrogen and two were fueled with a blend of CNG and hydrogen. The three blended-fuel vehicles were originally equipped with either factory CNG engines or factory gasoline engines that were converted to run CNG fuel. The vehicles were variously modified to operate on blended fuel and were tested using 15 to 50% blends of hydrogen (by volume). The pure-hydrogen-fueled vehicle was converted from gasoline fuel to operate on 100% hydrogen. All vehicles were fueled from the Arizona Public Service's Alternative Fuel Pilot Plant, which was developed to dispense gaseous fuels, including CNG, blends of CNG and hydrogen, and pure hydrogen with up to 99.9999% purity. The primary objective of the test was to evaluate the safety and reliability of operating vehicles on hydrogen and blended hydrogen fuel, and the interface between the vehicles and the hydrogen fueling infrastructure. A secondary objective was to quantify vehicle emissions, cost, and performance. Over a total of 40,000 fleet test miles, no safety issues were found. Also, significant reductions in emissions were achieved by adding hydrogen to the fuel. This report presents results of testing conducted over 6,864 kilometers (4,265 miles) of operation using the pure-hydrogen-fueled Mercedes Sprinter van.

  16. Advanced Vehicle Testing Activity: Hydrogen-Fueled Mercedes Sprinter Van -- Operating Summary

    SciTech Connect

    Karner, D.; Francfort, James Edward

    2003-01-01

    Over the past two years, Arizona Public Service, a subsidiary of Pinnacle West Capital Corporation, in cooperation with the U.S. Department of Energy's Advanced Vehicle Testing Activity, tested four gaseous fuel vehicles as part of its alternative fueled vehicle fleet. One vehicle operated initially using compressed natural gas (CNG) and later a blend of CNG and hydrogen. Of the other three vehicles, one was fueled with pure hydrogen and two were fueled with a blend of CNG and hydrogen. The three blended-fuel vehicles were originally equipped with either factory CNG engines or factory gasoline engines that were converted to run CNG fuel. The vehicles were variously modified to operate on blended fuel and were tested using 15 to 50% blends of hydrogen (by volume). The pure- hydrogen-fueled vehicle was converted from gasoline fuel to operate on 100% hydrogen. All vehicles were fueled from the Arizona Public Service's Alternative Fuel Pilot Plant, which was developed to dispense gaseous fuels, including CNG, blends of CNG and hydrogen, and pure hydrogen with up to 99.9999% purity. The primary objective of the test was to evaluate the safety and reliability of operating vehicles on hydrogen and blended hydrogen fuel, and the interface between the vehicles and the hydrogen fueling infrastructure. A secondary objective was to quantify vehicle emissions, cost, and performance. Over a total of 40,000 fleet test miles, no safety issues were found. Also, significant reductions in emissions were achieved by adding hydrogen to the fuel. This report presents results of testing conducted over 6,864 kilometers (4,265 miles) of operation using the pure-hydrogen-fueled Mercedes Sprinter van.

  17. 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.

  18. Advanced multiphysics coupling for LWR fuel performance analysis

    SciTech Connect

    Hales, J. D.; Tonks, M. R.; Gleicher, F. N.; Spencer, B. W.; Novascone, S. R.; Williamson, R. L.; Pastore, G.; Perez, D. M.

    2015-10-01

    Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics, particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is possible to use

  19. Advanced multiphysics coupling for LWR fuel performance analysis

    DOE PAGESBeta

    Hales, J. D.; Tonks, M. R.; Gleicher, F. N.; Spencer, B. W.; Novascone, S. R.; Williamson, R. L.; Pastore, G.; Perez, D. M.

    2015-10-01

    Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics,more » particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is

  20. 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.

  1. 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.

  2. Gasoline Ultra Efficient Fuel Vehicle with Advanced Low Temperature Combustion

    SciTech Connect

    Confer, Keith

    2014-09-30

    The objective of this program was to develop, implement and demonstrate fuel consumption reduction technologies which are focused on reduction of friction and parasitic losses and on the improvement of thermal efficiency from in-cylinder combustion. The program was executed in two phases. The conclusion of each phase was marked by an on-vehicle technology demonstration. Phase I concentrated on short term goals to achieve technologies to reduce friction and parasitic losses. The duration of Phase I was approximately two years and the target fuel economy improvement over the baseline was 20% for the Phase I demonstration. Phase II was focused on the development and demonstration of a breakthrough low temperature combustion process called Gasoline Direct- Injection Compression Ignition (GDCI). The duration of Phase II was approximately four years and the targeted fuel economy improvement was 35% over the baseline for the Phase II demonstration vehicle. The targeted tailpipe emissions for this demonstration were Tier 2 Bin 2 emissions standards.

  3. 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.

  4. A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility

    SciTech Connect

    S. Khericha

    2010-12-01

    The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3 to 5 years to construct.

  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. 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.

  7. Microwave Processing of Simulated Advanced Nuclear Fuel Pellets

    SciTech Connect

    D.E. Clark; D.C. Folz

    2010-08-29

    Throughout the three-year project funded by the Department of Energy (DOE) and lead by Virginia Tech (VT), project tasks were modified by consensus to fit the changing needs of the DOE with respect to developing new inert matrix fuel processing techniques. The focus throughout the project was on the use of microwave energy to sinter fully stabilized zirconia pellets using microwave energy and to evaluate the effectiveness of techniques that were developed. Additionally, the research team was to propose fundamental concepts as to processing radioactive fuels based on the effectiveness of the microwave process in sintering the simulated matrix material.

  8. Vehicle Data for Alternative Fuel Vehicles (AFVs) and Hybrid Fuel Vehicles (HEVs) from the Alternative Fuels and Advanced Vehicles Data Center (AFCD)

    DOE Data Explorer

    The AFDC provides search capabilities for many different models of both light-duty and heavy-duty vehicles. Engine and transmission type, fuel and class, fuel economy and emission certification are some of the facts available. The search will also help users locate dealers in their areas and do cost analyses. Information on alternative fuel vehicles and on advanced technology vehicles, along with calculators, resale and conversion information, links to incentives and programs such as Clean Cities, and dozens of fact sheets and publications make this section of the AFDC a valuable resource for car buyers.

  9. Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

    SciTech Connect

    Stauff, N.E.; Klim, T.K.; Taiwo, T.A.; Fiorina, C.; Franceschini, F.

    2013-07-01

    A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic

  10. Advanced diesel electronic fuel injection and turbocharging. Final report, July 1990-December 1993

    SciTech Connect

    Beck, N.J.; Barkhimer, R.L.; Steinmeyer, D.C.; Kelly, J.E.

    1993-12-01

    The program investigated advanced diesel air charging and fuel injection systems to improve specific power, fuel economy, noise, exhaust emissions, and cold startability. The techniques explored included variable fuel injection rate shaping, variable injection timing, full-authority electronic engine control, turbo-compound cooling, regenerative air circulation as a cold start aid, and variable geometry turbocharging. A Servojet electronic fuel injection system was designed and manufactured for the Cummins VTA-903 engine. A special Servojet twin turbocharger exhaust system was also installed. A series of high speed combustion flame photos was taken using the single cylinder optical engine at Michigan Technological University. Various fuel injection rate shapes and nozzle configurations were evaluated. Single-cylinder bench tests were performed to evaluate regenerative inlet air heating techniques as an aid to cold starting. An exhaust-driven axial cooling air fan was manufactured and tested on the VTA-903 engine. Electronic fuel injection, Turbocharging, Diesel combustion, Cold starting, Flame photography.

  11. Advanced coal-fueled industrial cogeneration gas turbine system

    SciTech Connect

    LeCren, R.T.; Cowell, L.H.; Galica, M.A.; Stephenson, M.D.; When, C.S.

    1992-06-01

    This report covers the activity during the period from 2 June 1991 to 1 June 1992. The major areas of work include: the combustor sub-scale and full size testing, cleanup, coal fuel specification and processing, the Hot End Simulation rig and design of the engine parts required for use with the coal-fueled combustor island. To date Solar has demonstrated: Stable and efficient combustion burning coal-water mixtures using the Two Stage Slagging Combustor; Molten slag removal of over 97% using the slagging primary and the particulate removal impact separator; and on-site preparation of CWM is feasible. During the past year the following tasks were completed: The feasibility of on-site CWM preparation was demonstrated on the subscale TSSC. A water-cooled impactor was evaluated on the subscale TSSC; three tests were completed on the full size TSSC, the last one incorporating the PRIS; a total of 27 hours of operation on CWM at design temperature were accumulated using candle filters supplied by Refraction through Industrial Pump Filter; a target fuel specification was established and a fuel cost model developed which can identify sensitivities of specification parameters; analyses of the effects of slag on refractory materials were conducted; and modifications continued on the Hot End Simulation Rig to allow extended test times.

  12. 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)

  13. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  14. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  15. s-Block Elements. Independent Learning Project for Advanced Chemistry (ILPAC). Unit I1.

    ERIC Educational Resources Information Center

    Inner London Education Authority (England).

    This unit is one of 10 first year units produced by the Independent Learning Project for Advanced Chemistry (ILPAC). The unit, which consists of two sections and an appendix, focuses on the elements and compounds of Groups I and II (the s-block) of the periodic table. The groups are treated concurrently to note comparisons between groups and to…

  16. Comparison of ash behavior of different fuels in fluidised bed combustion using advanced fuel analysis and global equilibrium calculations

    SciTech Connect

    Zevenhoven-Onderwater, M.; Blomquist, J.P.; Skrifvars, B.J.; Backman, R.; Hupa, M.

    1999-07-01

    The behavior of different ashes is predicted by means of a combination of an advanced fuel analysis and global equilibrium calculations. In order to cover a broad spectrum of fuels a coal, a peat, a forest residue and Salix (i.e. willow) are studied. The latter was taken with and without soil contamination, i.e. with a high and low content of silica , respectively. It is shown that mineral matter in fossil and biomass fuels can be present in the matrix of the fuel itself or as included minerals. Using an advanced fuel analysis, i.e. a fractionation method, this mineral content can be divided into four fractions. The first fraction mainly contains those metal ions, that can be leached out of the fuel by water and mainly contains alkali sulfates, carbonates and chlorides. The second fraction mainly consists of those ions leached out by ammonium acetate and covers those ions, that are connected to the organic matrix. The third fraction contains the metals leached out by hydrochloric acid and contains earth alkali carbonates and sulfates as well as pyrites. The rest fraction contains those minerals, that are not leached out by any of the above mentioned solvents, such as silicates. A global equilibrium analysis is used to predict the thermal and chemical behavior of the combined first and second fractions and of the combined third and rest fractions under pressurized and/or atmospheric combustion conditions. Results of both the fuel analysis and the global equilibrium analysis are discussed and practical implications for combustion processes are pointed out.

  17. Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

    SciTech Connect

    Jason Hales; Various

    2014-06-01

    The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

  18. Advanced Pellet-Cladding Interaction Modeling using the US DOE CASL Fuel Performance Code: Peregrine

    SciTech Connect

    Montgomery, Robert O.; Capps, Nathan A.; Sunderland, Dion J.; Liu, Wenfeng; Hales, Jason; Stanek, Chris; Wirth, Brian D.

    2014-06-15

    The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermo-mechanical-chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

  19. Recent Advances in Carbon Nanotube-Based Enzymatic Fuel Cells

    PubMed Central

    Cosnier, Serge; Holzinger, Michael; Le Goff, Alan

    2014-01-01

    This review summarizes recent trends in the field of enzymatic fuel cells. Thanks to the high specificity of enzymes, biofuel cells can generate electrical energy by oxidation of a targeted fuel (sugars, alcohols, or hydrogen) at the anode and reduction of oxidants (O2, H2O2) at the cathode in complex media. The combination of carbon nanotubes (CNT), enzymes and redox mediators was widely exploited to develop biofuel cells since the electrons involved in the bio-electrocatalytic processes can be efficiently transferred from or to an external circuit. Original approaches to construct electron transfer based CNT-bioelectrodes and impressive biofuel cell performances are reported as well as biomedical applications. PMID:25386555

  20. Recent advances in carbon nanotube-based enzymatic fuel cells.

    PubMed

    Cosnier, Serge; Holzinger, Michael; Le Goff, Alan

    2014-01-01

    This review summarizes recent trends in the field of enzymatic fuel cells. Thanks to the high specificity of enzymes, biofuel cells can generate electrical energy by oxidation of a targeted fuel (sugars, alcohols, or hydrogen) at the anode and reduction of oxidants (O2, H2O2) at the cathode in complex media. The combination of carbon nanotubes (CNT), enzymes and redox mediators was widely exploited to develop biofuel cells since the electrons involved in the bio-electrocatalytic processes can be efficiently transferred from or to an external circuit. Original approaches to construct electron transfer based CNT-bioelectrodes and impressive biofuel cell performances are reported as well as biomedical applications. PMID:25386555

  1. Interatomic potentials for mixed oxide and advanced nuclear fuels

    SciTech Connect

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

    2011-03-01

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

  2. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during

  3. 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.

  4. Advanced fuel cell development. Progress report, April-June 1984

    SciTech Connect

    Pierce, R.D.; Claar, T.D.; Dees, D.W.; Fousek, R.J.; Kaun, T.D.; Kucera, G.H.; Minh, N.Q.; Mrazek, F.C.; Poeppel, R.B.; Smith, J.L.

    1984-11-01

    This report describes fuel cell research and development activities at Argonne National Laboratory (ANL) during the period April through June 1984. These efforts have been directed toward seeking alternative cathode materials to NiO for molten carbonate fuel cells. Particular emphasis has been placed on studying the relationship between synthesis conditions and the resistivity of doped and undoped LiFeO/sub 2/ and Li/sub 2/MnO/sub 3/ and on achieving a better understanding of the crystalline defect structures of the thermodynamically stable phases. To this end, several experimental assemblies (including synthesis, solubility, and sintering vessels and a high-pressure thermogravimetric analyzer) have been constructed to permit 10-atm operation. In addition, data on solubility and cathode cation-LiAlO/sub 2/ interaction were taken for NiO, Li/sub 2/MnO/sub 3/, Mg-doped Li/sub 2/MnO/sub 3/, LiFeO/sub 2/, and ZnO at 1- and 10-atm pressure, and ion migration from LiFeO/sub 2/ and Li/sub 2/MnO/sub 3/ in the cell environment for 200 and 1000 h was examined. Techniques are being studied for the preparation of thin electrode and electrolyte materials by tape casting. A study to provide improved understanding of anode creep and densification occurring under fuel cell conditions is under way.

  5. Thermal management of advanced fuel cell power systems

    NASA Technical Reports Server (NTRS)

    Vanderborgh, N. E.; Hedstrom, J.; Huff, J.

    1990-01-01

    It is shown that fuel cell devices are particularly attractive for the high-efficiency, high-reliability space hardware necessary to support upcoming space missions. These low-temperature hydrogen-oxygen systems necessarily operate with two-phase water. In either PEMFCs (proton exchange membrane fuel cells) or AFCs (alkaline fuel cells), engineering design must be critically focused on both stack temperature control and on the relative humidity control necessary to sustain appropriate conductivity within the ionic conductor. Water must also be removed promptly from the hardware. Present designs for AFC space hardware accomplish thermal management through two coupled cooling loops, both driven by a heat transfer fluid, and involve a recirculation fan to remove water and heat from the stack. There appears to be a certain advantage in using product water for these purposes within PEM hardware, because in that case a single fluid can serve both to control stack temperature, operating simultaneously as a heat transfer medium and through evaporation, and to provide the gas-phase moisture levels necessary to set the ionic conductor at appropriate performance levels. Moreover, the humidification cooling process automatically follows current loads. This design may remove the necessity for recirculation gas fans, thus demonstrating the long-term reliability essential for future space power hardware.

  6. 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.

  7. Engineering development of advanced physical fine coal cleaning for premium fuel applications. Quarterly technical progress report No. 1, October--December 1992

    SciTech Connect

    Smit, F.J.; Jha, M.C.

    1993-01-18

    This project is a step in the Department of Energy`s program to show that ultra-clean fuel can be produced from selected coals and that the fuel will be a cost-effective replacement for oil and natural gas now fueling boilers in this country. The replacement of premium fossil fuels with coal can only be realized if retrofit costs are kept to a minimum and retrofit boiler emissions meet national goals for clean air. These concerns establish the specifications for maximum ash and sulfur levels and combustion properties of the ultra-clean coal. The primary objective is to develop the design base for prototype commercial advanced fine coal cleaning facilities capable of producing ultra-clean coals suitable for conversion to coal-water slurry fuel. The fine coal cleaning technologies are advanced column flotation and selective agglomeration. A secondary objective is to develop the design base for near-term commercial integration of advanced fine coal cleaning technologies in new or existing coal preparation plants for economically and efficiently processing minus 28-mesh coal fines. A third objective is to determine the distribution of toxic trace elements between clean coal and refuse when applying the advance column flotation and selective agglomeration technologies. The project team consists of Amax Research & Development Center (Amax R&D), Amax Coal industries, Bechtel Corporation, Center for Applied Energy Research (CAER) at the University of Kentucky, and Arcanum Corporation.

  8. Engineereing development of advanced physical fine coal cleaning for premium fuel applications. Quarterly technical progress report No. 5, October--December 1993

    SciTech Connect

    Smit, F.J.; Jha, M.C.

    1994-02-18

    This project is a major step in the Department of Energy`s program to show that ultra-clean coal-water slurry fuel (CWF) can be produced from selected coals and that this premium fuel will be a cost-effective replacement for oil and natural gas now fueling some of the industrial and utility boilers in the United States. The replacement of oil and gas with CWF can only be realized if retrofit costs are kept to a minimum and retrofit boiler emissions meet national goals for clean air. These concerns establish the specifications for maximum ash and sulfur levels and combustion properties of the CWF. The project has three major objectives: The primary objective is to develop the design base for prototype commercial advanced fine coal cleaning facilities capable of producing ultra-clean coals suitable for conversion to coal-water slurry fuel for premium fuel applications. The fine coal cleaning technologies are advanced column flotation and selective agglomeration. A secondary objective is to develop the design base for near-term application of these advanced fine coal cleaning technologies in new or existing coal preparation plants for efficiently processing minus 28-mesh coal fines and converting this to marketable products in current market economics. A third objective is to determine the removal of toxic trace elements from coal by advance column flotation and selective agglomeration technologies.

  9. Safeguards and Non-proliferation Issues as Related to Advanced Fuel Cycle and Advanced Fast Reactor Development with Processing of Reactor Fuel

    SciTech Connect

    Rahmat Aryaeinejad; Jerry D. Cole; Mark W. Drigert; Dee E. Vaden

    2006-10-01

    The goal of this work is to establish basic data and techniques to enable safeguards appropriate to a new generation of nuclear power systems that will be based on fast spectrum reactors and mixed actinide fuels containing significant quantities of "minor" actinides, possibly due to reprocessing, and determination of what new radiation signatures and parameters need to be considered. The research effort focuses on several problems associated with the use of fuel having significantly different actinide inventories that current practice and on the development of innovative techniques using new radiation signatures and other parameters useful for safeguards and monitoring. In addition, the development of new distinctive radiation signatures as an aid in controlling proliferation of nuclear materials has parallel applications to support Gen-IV and current advanced fuel cycle initiative (AFCI) goals as well as the anticipated Global Nuclear Energy Partnership (GNEP).

  10. Engineering development of advance physical fine coal cleaning for premium fuel applications

    SciTech Connect

    Jha, M.C.; Smit, F.J.; Shields, G.L.

    1995-11-01

    The objective of this project is to develop the engineering design base for prototype fine coal cleaning plants based on Advanced Column Flotation and Selective Agglomeration processes for premium fuel and near-term applications. Removal of toxic trace elements is also being investigated. The scope of the project includes laboratory research and bench-scale testing of each process on six coals followed by design, construction, and operation of a 2 tons/hour process development unit (PDU). Three coals will be cleaned in tonnage quantity and provided to DOE and its contractors for combustion evaluation. Amax R&D (now a subsidiary of Cyprus Amax Mineral Company) is the prime contractor. Entech Global is managing the project and performing most of the research and development work as an on-site subcontractor. Other participants in the project are Cyprus Amax Coal Company, Arcanum, Bechtel, TIC, University of Kentucky and Virginia Tech. Drs. Keller of Syracuse and Dooher of Adelphi University are consultants.

  11. 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.

  12. PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

    SciTech Connect

    S. T. Khericha

    2007-04-01

    The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.

  13. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  14. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    PubMed

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR. PMID:19776247

  15. Advanced thermally stable jet fuels. Technical progress report, July 1995--September 1995

    SciTech Connect

    Schobert, H.H.; Eser, S.; Song, C.

    1995-10-01

    The Penn State program in advanced thermally stable jet engine fuels has five components: development of mechanisms of degradation and solids formation; quantitative measurement of growth of sub-micrometer-sized and micrometer particles suspended in fuels during thermal stresses; characterization of carbonaceous deposits by various instrumental and microscopic methods; elucidation of the role of additives in retarding the formation of carbonaceous solids; and assessment of the potential of producing high yields of cycloalkanes and hydroaromatics by direct coal liquefaction. Progress is described.

  16. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

    NASA Astrophysics Data System (ADS)

    Terrani, K. A.; Zinkle, S. J.; Snead, L. L.

    2014-05-01

    Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.

  17. Global shielding analysis for the three-element core advanced neutron source reactor under normal operating conditions

    SciTech Connect

    Slater, C.O.; Bucholz, J.A.

    1995-08-01

    Two-dimensional discrete ordinates radiation transport calculations were performed for a model of the three-element core Advanced Neutron Source reactor design under normal operating conditions. The core consists of two concentric upper elements and a lower element radially centered in the annulus between the upper elements. The initial radiation transport calculations were performed with the DORT two-dimensional discrete ordinates radiation transport code using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub 6} quadrature, and a P{sub 1} Legendre polynomial expansion of the cross sections to determine the fission neutron source distribution in the core fuel elements. These calculations were limited to neutron groups only. The final radiation transport calculations, also performed with DORT using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub l0} quadrature, and a P{sub 3} Legendre polynomial expansion of the cross sections, produced neutron and gamma-ray fluxes over the full extent of the geometry model. Responses (or activities) at various locations in the model were then obtained by folding the appropriate response functions with the fluxes at those locations. Some comparisons were made with VENTURE-calculated (diffusion theory) 20-group neutron fluxes that were summed into four broad groups. Tne results were in reasonably good agreement when the effects of photoneutrons were not included, thus verifying the physics model upon which the shielding model was based. Photoneutrons increased the fast-neutron flux levels deep within the D{sub 2}0 several orders of magnitude. Results are presented as tables of activity values for selected radial and axial traverses, plots of the radial and axial traverse data, and activity contours superimposed on the calculational geometry model.

  18. 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.

  19. Evaluation of advanced lift concepts and potential fuel conservation for short-haul aircraft

    NASA Technical Reports Server (NTRS)

    Sweet, H. S.; Renshaw, J. H.; Bowden, M. K.

    1975-01-01

    The effect of different field lengths, cruise requirements, noise level, and engine cycle characteristics on minimizing fuel consumption and minimizing operating cost at high fuel prices were evaluated for some advanced short-haul aircraft. The conceptual aircraft were designed for 148 passengers using the upper surface-internally blown jet flap, the augmentor wing, and the mechanical flap lift systems. Advanced conceptual STOL engines were evaluated as well as a near-term turbofan and turboprop engine. Emphasis was given to designs meeting noise levels equivalent to 95-100 EPNdB at 152 m (500 ft) sideline.

  20. Advanced Fuel Cycle Initiative - Projected Linear Heat Generation Rate and Burnup Calculations

    SciTech Connect

    Richard G. Ambrosek; Gray S. Chang; Debbie J. Utterbeck

    2005-02-01

    This report provides documentation of the physics analysis performed to determine the linear heat generation rate (LHGR) and burnup calculations for the Advanced Fuel Cycle Initiative (AFCI) tests, AFC-1D, AFC-1H, and AFC-1G. The AFC-1D and AFC-1H tests consists of low-fertile metallic fuel compositions and the AFC-1G test consists of non-fertile and low-fertile nitride compositions. These tests will be irradiated in the East Flux Trap (EFT) positions E1, E2, and E3, respectively, during Advanced Test Reactor (ATR) Cycle 135B.

  1. Recovery of Information from the Fast Flux Test Facility for the Advanced Fuel Cycle Initiative

    SciTech Connect

    Nielsen, Deborah L.; Makenas, Bruce J.; Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.

    2009-09-30

    The Fast Flux Test Facility is the most recent Liquid Metal Reactor to operate in the United States. Information from the design, construction, and operation of this reactor was at risk as the facilities associated with the reactor are being shut down. The Advanced Fuel Cycle Initiative is a program managed by the Office of Nuclear Energy of the U.S. Department of Energy with a mission to develop new fuel cycle technologies to support both current and advanced reactors. Securing and preserving the knowledge gained from operation and testing in the Fast Flux Test Facility is an important part of the Knowledge Preservation activity in this program.

  2. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    SciTech Connect

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program.

  3. 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.

  4. 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.

  5. Advanced Automotive Fuels Research, Development, and Commercialization Cluster (OH)

    SciTech Connect

    Linkous, Clovis; Hripko, Michael; Abraham, Martin; Balendiran, Ganesaratnam; Hunter, Allen; Lovelace-Cameron, Sherri; Mette, Howard; Price, Douglas; Walker, Gary; Wang, Ruigang

    2013-08-31

    Technical aspects of producing alternative fuels that may eventually supplement or replace conventional the petroleum-derived fuels that are presently used in vehicular transportation have been investigated. The work was centered around three projects: 1) deriving butanol as a fuel additive from bacterial action on sugars produced from decomposition of aqueous suspensions of wood cellulose under elevated temperature and pressure; 2) using highly ordered, openly structured molecules known as metal-organic framework (MOF) compounds as adsorbents for gas separations in fuel processing operations; and 3) developing a photocatalytic membrane for solar-driven water decomposition to generate pure hydrogen fuel. Several departments within the STEM College at YSU contributed to the effort: Chemistry, Biology, and Chemical Engineering. In the butanol project, sawdust was blended with water at variable pH and temperature (150 – 250{degrees}C), and heated inside a pressure vessel for specified periods of time. Analysis of the extracts showed a wide variety of compounds, including simple sugars that bacteria are known to thrive upon. Samples of the cellulose hydrolysate were fed to colonies of Clostridium beijerinckii, which are known to convert sugars to a mixture of compounds, principally butanol. While the bacteria were active toward additions of pure sugar solutions, the cellulose extract appeared to inhibit butanol production, and furthermore encouraged the Clostridium to become dormant. Proteomic analysis showed that the bacteria had changed their genetic code to where it was becoming sporulated, i.e., the bacteria were trying to go dormant. This finding may be an opportunity, as it may be possible to genetically engineer bacteria that resist the butanol-driven triggering mechanism to stop further fuel production. Another way of handling the cellulosic hydrolysates was to simply add the enzymes responsible for butanol synthesis to the hydrolytic extract ex-vivo. These

  6. Fuel Effects on Ignition and Their Impact on Advanced Combustion Engines (Poster)

    SciTech Connect

    Taylor, J.; Li, H.; Neill, S.

    2006-08-01

    The objective of this report is to develop a pathway to use easily measured ignition properties as metrics for characterizing fuels in advanced combustion engine research--correlate IQT{trademark} measured parameters with engine data. In HCCL engines, ignition timing depends on the reaction rates throughout compression stroke: need to understand sensitivity to T, P, and [O{sub 2}]; need to rank fuels based on more than one set of conditions; and need to understand how fuel composition (molecular species) affect ignition properties.

  7. Evaluation of advanced lift concepts and fuel conservative short-haul aircraft, volume 1

    NASA Technical Reports Server (NTRS)

    Renshaw, J. H.; Bowden, M. K.; Narucki, C. W.; Bennett, J. A.; Smith, P. R.; Ferrill, R. S.; Randall, C. C.; Tibbetts, J. G.; Patterson, R. W.; Meyer, R. T.

    1974-01-01

    The performance and economics of a twin-engine augmentor wing airplane were evaluated in two phases. Design aspects of the over-the-wing/internally blown flap hybrid, augmentor wing, and mechanical flap aircraft were investigated for 910 m. field length with parametric extension to other field lengths. Fuel savings achievable by application of advanced lift concepts to short-haul aircraft were evaluated and the effect of different field lengths, cruise requirements, and noise levels on fuel consumption and airplane economics at higher fuel prices were determined. Conclusions and recommendations are presented.

  8. Electrocatalyst advances for hydrogen oxidation in phosphoric acid fuel cells

    NASA Technical Reports Server (NTRS)

    Stonehart, P.

    1984-01-01

    The important considerations that presently exist for achieving commercial acceptance of fuel cells are centered on cost (which translates to efficiency) and lifetime. This paper addresses the questions of electrocatalyst utilization within porous electrode structures and the preparation of low-cost noble metal electrocatalyst combinations with extreme dispersions of the metal. Now that electrocatalyst particles can be prepared with dimensions of 10 A, either singly or in alloy combinations, a very large percentage of the noble metal atoms in a crystallite are available for reaction. The cost savings for such electrocatalysts in the present commercially driven environment are considerable.

  9. Advanced Method to Estimate Fuel Slosh Simulation Parameters

    NASA Technical Reports Server (NTRS)

    Schlee, Keith; Gangadharan, Sathya; Ristow, James; Sudermann, James; Walker, Charles; Hubert, Carl

    2005-01-01

    The nutation (wobble) of a spinning spacecraft in the presence of energy dissipation is a well-known problem in dynamics and is of particular concern for space missions. The nutation of a spacecraft spinning about its minor axis typically grows exponentially and the rate of growth is characterized by the Nutation Time Constant (NTC). For launch vehicles using spin-stabilized upper stages, fuel slosh in the spacecraft propellant tanks is usually the primary source of energy dissipation. For analytical prediction of the NTC this fuel slosh is commonly modeled using simple mechanical analogies such as pendulums or rigid rotors coupled to the spacecraft. Identifying model parameter values which adequately represent the sloshing dynamics is the most important step in obtaining an accurate NTC estimate. Analytic determination of the slosh model parameters has met with mixed success and is made even more difficult by the introduction of propellant management devices and elastomeric diaphragms. By subjecting full-sized fuel tanks with actual flight fuel loads to motion similar to that experienced in flight and measuring the forces experienced by the tanks these parameters can be determined experimentally. Currently, the identification of the model parameters is a laborious trial-and-error process in which the equations of motion for the mechanical analog are hand-derived, evaluated, and their results are compared with the experimental results. The proposed research is an effort to automate the process of identifying the parameters of the slosh model using a MATLAB/SimMechanics-based computer simulation of the experimental setup. Different parameter estimation and optimization approaches are evaluated and compared in order to arrive at a reliable and effective parameter identification process. To evaluate each parameter identification approach, a simple one-degree-of-freedom pendulum experiment is constructed and motion is induced using an electric motor. By applying the

  10. Statistical Methods Handbook for Advanced Gas Reactor Fuel Materials

    SciTech Connect

    J. J. Einerson

    2005-05-01

    Fuel materials such as kernels, coated particles, and compacts are being manufactured for experiments simulating service in the next generation of high temperature gas reactors. These must meet predefined acceptance specifications. Many tests are performed for quality assurance, and many of these correspond to criteria that must be met with specified confidence, based on random samples. This report describes the statistical methods to be used. The properties of the tests are discussed, including the risk of false acceptance, the risk of false rejection, and the assumption of normality. Methods for calculating sample sizes are also described.

  11. SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: Fourth Results Report

    SciTech Connect

    Eudy, L.; Chandler, K.

    2013-01-01

    SunLine Transit Agency, which provides public transit services to the Coachella Valley area of California, has demonstrated hydrogen and fuel cell bus technologies for more than 10 years. In May 2010, SunLine began demonstrating the advanced technology (AT) fuel cell bus with a hybrid electric propulsion system, fuel cell power system, and lithium-based hybrid batteries. This report describes operations at SunLine for the AT fuel cell bus and five compressed natural gas buses. The U.S. Department of Energy's National Renewable Energy Laboratory (NREL) is working with SunLine to evaluate the bus in real-world service to document the results and help determine the progress toward technology readiness. NREL has previously published three reports documenting the operation of the fuel cell bus in service. This report provides a summary of the results with a focus on the bus operation from February 2012 through November 2012.

  12. Technical basis for extending storage of the UK's advanced gas-cooled reactor fuel

    SciTech Connect

    Hambley, D.I.

    2013-07-01

    The UK Nuclear Decommissioning Agency has recently declared a date for cessation of reprocessing of oxide fuel from the UK's Advanced Gas-cooled Reactors (AGRs). This will fundamentally change the management of AGR fuel: from short term storage followed by reprocessing to long term fuel storage followed, in all likelihood, by geological disposal. In terms of infrastructure, the UK has an existing, modern wet storage asset that can be adapted for centralised long term storage of dismantled AGR fuel under the required pond water chemistry. No AGR dry stores exist, although small quantities of fuel have been stored dry as part of experimental programmes in the past. These experimental programmes have shown concerns about corrosion rates.

  13. Advanced NaBH4/H2O2 Fuel Cell for Space Applications

    NASA Astrophysics Data System (ADS)

    Miley, George H.; Kim, Kyu-Jung; Luo, Nie; Shrestha, Prajakti Joshi

    2009-03-01

    Fuel cells have played an important role in NASA's space program starting with the Gemini space program. However, improved fuel cell performance will be needed to enable demanding future missions. An advanced fuel cell (FC) using liquid fuel and oxidizer is being developed by U of IL/NPL team to provide air independence and to achieve higher power densities than normal H2/O2 fuel cells (Lou et al., 2008; Miley, 2007). Hydrogen peroxide (H2O2) is used in this FC directly at the cathode (Lou and Miley, 2004). Either of two types of reactant, namely a gas-phase hydrogen or an aqueous NaBH4 solution, is utilized as fuel at the anode. Experiments with both 10-W single cells and 500-W stacks demonstrate that the direct utilization of H2O2 and NaBH4 at the electrodes result in >30% higher voltage output compared to the ordinary H2/O2 FC (Miley, 2007). Further, the use of this combination of all liquid fuels provides—from an operational point of view—significant advantages (ease of storage, reduced pumping requirements, simplified heat removal). This design is inherently compact compared to other fuel cells that use gas phase reactants. This results in a high overall system (including fuel tanks, pumps and piping, waste heat radiator) power density. Further, work is in progress on a regenerative version which uses an electrical input, e.g. from power lines or a solar panel to regenerate reactants.

  14. Finite element method for optimal guidance of an advanced launch vehicle

    NASA Technical Reports Server (NTRS)

    Hodges, Dewey H.; Bless, Robert R.; Calise, Anthony J.; Leung, Martin

    1992-01-01

    A temporal finite element based on a mixed form of Hamilton's weak principle is summarized for optimal control problems. The resulting weak Hamiltonian finite element method is extended to allow for discontinuities in the states and/or discontinuities in the system equations. An extension of the formulation to allow for control inequality constraints is also presented. The formulation does not require element quadrature, and it produces a sparse system of nonlinear algebraic equations. To evaluate its feasibility for real-time guidance applications, this approach is applied to the trajectory optimization of a four-state, two-stage model with inequality constraints for an advanced launch vehicle. Numerical results for this model are presented and compared to results from a multiple-shooting code. The results show the accuracy and computational efficiency of the finite element method.

  15. 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.

  16. 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.

  17. ADVANCED HYDROGEN TRANSPORT MEMBRANES FOR VISION 21 FOSSIL FUEL PLANTS

    SciTech Connect

    Shane E. Roark; Tony F. Sammells; Adam Calihman; Andy Girard; Pamela M. Van Calcar; Richard Mackay; Tom Barton; Sara Rolfe

    2001-01-30

    Eltron Research Inc., and team members CoorsTek, McDermott Technology, Inc., Sued Chemie, Argonne National Laboratory, and Oak Ridge National Laboratory are developing an environmentally benign, inexpensive, and efficient method for separating hydrogen from gas mixtures produced during industrial processes, such as coal gasification. This objective is being pursued using dense membranes based in part on Eltron-patented ceramic materials with a demonstrated ability for proton and electron conduction. The technical goals are being addressed by modifying single-phase and composite membrane composition and microstructure to maximize proton and electron conductivity without loss of material stability. Ultimately, these materials must enable hydrogen separation at practical rates under ambient and high-pressure conditions, without deactivation in the presence of feedstream components such as carbon dioxide, water, and sulfur. This project was motivated by the National Energy Technology Laboratory (NETL) Vision 21 initiative which seeks to economically eliminate environmental concerns associated with the use of fossil fuels. The proposed technology addresses the DOE Vision 21 initiative in two ways. First, this process offers a relatively inexpensive solution for pure hydrogen separation that can be easily incorporated into Vision 21 fossil fuel plants. Second, this process could reduce the cost of hydrogen, which is a clean burning fuel under increasing demand as supporting technologies are developed for hydrogen utilization and storage. Additional motivation for this project arises from the potential of this technology for other applications. Membranes testing during this reporting period were greater than 1 mm thick and had the general perovskite composition AB{sub 1-x}B'{sub x}O{sub 3-{delta}}, where 0.05 {<=} x {<=} 0.3. These materials demonstrated hydrogen separation rates between 1 and 2 mL/min/cm{sup 2}, which represents roughly 20% of the target goal for

  18. Use of freeze-casting in advanced burner reactor fuel design

    SciTech Connect

    Lang, A. L.; Yablinsky, C. A.; Allen, T. R.; Burger, J.; Hunger, P. M.; Wegst, U. G. K.

    2012-07-01

    This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary

  19. Engineering metabolic systems for production of advanced fuels.

    PubMed

    Yan, Yajun; Liao, James C

    2009-04-01

    The depleting petroleum storage and increasing environmental deterioration are threatening the sustainable development of human societies. As such, biofuels and chemical feedstocks generated from renewable sources are becoming increasingly important. Although previous efforts led to great success in bio-ethanol production, higher alcohols, fatty acid derivatives including biodiesels, alkanes, and alkenes offer additional advantages because of their compatibility with existing infrastructure. In addition, some of these compounds are useful chemical feedstocks. Since native organisms do not naturally produce these compounds in high quantities, metabolic engineering becomes essential in constructing producing organisms. In this article, we briefly review the four major metabolic systems, the coenzyme-A mediated pathways, the keto acid pathways, the fatty acid pathway, and the isoprenoid pathways, that allow production of these fuel-grade chemicals. PMID:19198907

  20. Advanced Materials for PEM-Based Fuel Cell Systems

    SciTech Connect

    James E. McGrath; Donald G. Baird; Michael von Spakovsky

    2005-10-26

    Proton exchange membrane fuel cells (PEMFCs) are quickly becoming attractive alternative energy sources for transportation, stationary power, and small electronics due to the increasing cost and environmental hazards of traditional fossil fuels. Two main classes of PEMFCs include hydrogen/air or hydrogen/oxygen fuel cells and direct methanol fuel cells (DMFCs). The current benchmark membrane for both types of PEMFCs is Nafion, a perfluorinated sulfonated copolymer made by DuPont. Nafion copolymers exhibit good thermal and chemical stability, as well as very high proton conductivity under hydrated conditions at temperatures below 80 degrees C. However, application of these membranes is limited due to their high methanol permeability and loss of conductivity at high temperatures and low relative humidities. These deficiencies have led to the search for improved materials for proton exchange membranes. Potential PEMs should have good thermal, hydrolytic, and oxidative stability, high proton conductivity, selective permeability, and mechanical durability over long periods of time. Poly(arylene ether)s, polyimides, polybenzimidazoles, and polyphenylenes are among the most widely investigated candidates for PEMs. Poly(arylene ether)s are a promising class of proton exchange membranes due to their excellent thermal and chemical stability and high glass transition temperatures. High proton conductivity can be achieved through post-sulfonation of poly(arylene ether) materials, but this most often results in very high water sorption or even water solubility. Our research has shown that directly polymerized poly(arylene ether) copolymers show important advantages over traditional post-sulfonated systems and also address the concerns with Nafion membranes. These properties were evaluated and correlated with morphology, structure-property relationships, and states of water in the membranes. Further improvements in properties were achieved through incorporation of inorganic