Science.gov

Sample records for fuel element behavior

  1. 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)

  2. Advancements in the behavioral modeling of fuel elements and related structures

    SciTech Connect

    Billone, M.C.; Montgomery, R.O.; Rashid, Y.R.; Head, J.L.; ANATECH Research Corp., San Diego, CA; Royal Naval Coll., Greenwich )

    1989-01-01

    An important aspect of the design and analysis of nuclear reactors is the ability to predict the behavior of fuel elements in the adverse environment of a reactor system. By understanding the thermomechanical behavior of the different materials which constitute a nuclear fuel element, analysis and predictions can be made regarding the integrity and reliability of fuel element designs. The SMiRT conference series, through the division on fuel elements and the post-conference seminars on fuel element modeling, provided technical forums for the international participation in the exchange of knowledge concerning the thermomechanical modeling of fuel elements. This paper discusses the technical advances in the behavioral modeling of fuel elements presented at the SMiRT conference series since its inception in 1971. Progress in the areas of material properties and constitutive relationships, modeling methodologies, and integral modeling approaches was reviewed and is summarized in light of their impact on the thermomechanical modeling of nuclear fuel elements. 34 refs., 5 tabs.

  3. FUEL ELEMENT

    DOEpatents

    Fortescue, P.; Zumwalt, L.R.

    1961-11-28

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

  4. FUEL ELEMENT

    DOEpatents

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

    1962-03-01

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

  5. Vibration behavior of fuel-element vibration suppressors for the advanced power reactor

    NASA Technical Reports Server (NTRS)

    Adams, D. W.; Fiero, I. B.

    1973-01-01

    Preliminary shock and vibration tests were performed on vibration suppressors for the advanced power reactor for space application. These suppressors position the fuel pellets in a pin type fuel element. The test determined the effect of varying axial clearance on the behavior of the suppressors when subjected to shock and vibratory loading. The full-size suppressor was tested in a mockup model of fuel and clad which required scaling of test conditions. The test data were correlated with theoretical predictions for suppressor failure. Good agreement was obtained. The maximum difference with damping neglected was about 30 percent. Neglecting damping would result in a conservative design.

  6. Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

    NASA Astrophysics Data System (ADS)

    Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

    2010-04-01

    The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

  7. 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.

  8. 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

  9. 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)

  10. 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.

  11. 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.

  12. 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.

  13. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Picklesimer, M.L.; Thurber, W.C.

    1961-01-01

    A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

  14. 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.

  15. CONCENTEIC TUBULAR FUEL ELEMENT

    DOEpatents

    Wheelock, C.W.

    1960-08-16

    An improved fuel element for an organic-moderated reactor was designed that comprises an inner and an outer container tube, a plurality of spaced, concentric fuel tubes positioned between the container tubes, each of the fuel tubes comprising a core of fissionable material with cladding on the sides thereof, each of the sides having a plurality of fins, the fuel tubes and the container tubes defining annular spaces for coolant flow, and the inner container tube defining a channel for a reactor moderator.

  16. 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)

  17. 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.

  18. 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)

  19. 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)

  20. 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.

  1. 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.

  2. TWISTED RIBBON FUEL ELEMENT

    DOEpatents

    Breden, C.R.; Schultz, A.B.

    1961-06-01

    A reactor core formed of bundles of parallel fuel elements in the form of ribbons is patented. The fuel ribbons are twisted about their axes so as to have contact with one another at regions spaced lengthwise of the ribbons and to be out of contact with one another at locations between these spaced regions. The contact between the ribbons is sufficient to allow them to be held together in a stable bundle in a containing tube without intermediate support, while permitting enough space between the ribbon for coolant flowing.

  3. 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.

  4. 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)

  5. 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)

  6. 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.

  7. 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.

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  12. 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

  13. 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)

  14. 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.

  15. 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)

  16. 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)

  17. 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)

  18. 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.

  19. 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)

  20. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

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

    1959-10-01

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

  1. 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)

  2. Dryout of BWR fuel elements

    SciTech Connect

    Reisch, Frigyes

    2006-07-01

    To increase the power output of the presently operating power reactors is a worldwide trend. One limiting factor from the safety and commercial point of views is the maximum allowable thermal load of the fuel. The findings of the presented loop experiments are that the margin to the burnout of the fuel elements can be defined by a single parameter the void. (authors)

  3. 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.

  4. 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)

  5. 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.

  6. 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.

  7. 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.

  8. FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Carney, K.G. Jr.

    1959-07-14

    A nuclear fuel element comprising a large number og wafers of fissionable material and a protective jacket having compartments holding these wafers is described. The compartments of the jacket aid the removal of heat from the wafers, keep the wafers or fragments thereof from migrating in the jacket, and permit the escape of gaseous fission products.

  9. 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.

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

    SciTech Connect

    Hai Huang; Ben Spencer; Jason Hales

    2014-10-01

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

  11. 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.

  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. 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.

  14. FUEL ELEMENT FOR A NEUTRONIC REACTOR

    DOEpatents

    McGeary, R.K.; Winslow, F.R.

    1963-08-13

    A method of making fuel elements wherein several individual fuel pellets are positioned into a cladding tube and the tape stretched longitudinally until the cladding tube grips each pellet and, in addition, necks down between each pellet is described. (AEC)

  15. Automated Fuel Element Closure Welding System

    SciTech Connect

    Wahlquist, D.R.

    1993-01-01

    The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout.

  16. Automated Fuel Element Closure Welding System

    SciTech Connect

    Wahlquist, D.R.

    1993-03-01

    The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout.

  17. 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.

  18. 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)

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. HTGR fuel element structural design considerations

    SciTech Connect

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1986-09-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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.

  17. 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)

  18. 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)

  19. 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.

  20. 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)

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. Spring element for holding down nuclear reactor fuel assembly

    SciTech Connect

    Steinke, A.

    1981-07-14

    Spring element is described for holding down and bracing a fuel assembly against a hold-down plate upwardly limiting the reactor core of a nuclear reactor. Includes a spring-loaded rod-shaped member separately formed independently of the fuel assembly and being slidable axially and form-lockingly into the fuel assembly.

  6. 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.

  7. Multidimensional multiphysics simulation of nuclear fuel behavior

    NASA Astrophysics Data System (ADS)

    Williamson, R. L.; Hales, J. D.; Novascone, S. R.; Tonks, M. R.; Gaston, D. R.; Permann, C. J.; Andrs, D.; Martineau, R. C.

    2012-04-01

    Nuclear fuel operates in an environment that induces complex multiphysics phenomena, occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. This multiphysics behavior is often tightly coupled and many important aspects are inherently multidimensional. Most current fuel modeling codes employ loose multiphysics coupling and are restricted to 2D axisymmetric or 1.5D approximations. This paper describes a new modeling tool able to simulate coupled multiphysics and multiscale fuel behavior, for either 2D axisymmetric or 3D geometries. Specific fuel analysis capabilities currently implemented in this tool are described, followed by a set of demonstration problems which include a 10-pellet light water reactor fuel rodlet, three-dimensional analysis of pellet clad mechanical interaction in the vicinity of a defective fuel pellet, coupled heat transfer and fission product diffusion in a TRISO-coated fuel particle, a demonstration of the ability to couple to lower-length scale models to account for material property variation with microstructural evolution, and a demonstration of the tool's ability to efficiently solve very large and complex problems using massively-parallel computing. A final section describes an early validation exercise, comparing simulation results to a light water reactor fuel rod experiment.

  8. Nuclear fuel elements and method of making same

    DOEpatents

    Schweitzer, Donald G.

    1992-01-01

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

  9. NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

    DOEpatents

    Brooks, H.

    1960-04-26

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

  10. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    SciTech Connect

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

    2011-10-01

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

  11. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOEpatents

    Flint, O.

    1961-01-10

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

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

    SciTech Connect

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

    1995-12-31

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

  13. Environmental mineralogy - Understanding element behavior in ecosystems

    NASA Astrophysics Data System (ADS)

    Brown, Gordon E., Jr.; Calas, Georges

    2011-02-01

    Environmental Mineralogy has developed over the past decade in response to the recognition that minerals are linked in many important ways with the global ecosystem. Minerals are the main repositories of the chemical elements in Earth's crust and thus are the main sources of elements needed for the development of civilization, contaminant and pollutant elements that impact global and local ecosystems, and elements that are essential plant nutrients. These elements are released from minerals through natural processes, such as chemical weathering, and anthropogenic activities, such as mining and energy production, agriculture and industrial activities, and careless waste disposal. Minerals also play key roles in the biogeochemical cycling of the elements, sequestering elements and releasing them as the primary minerals in crustal rocks undergo various structural and compositional transformations in response to physical, chemical, and biological processes that produce secondary minerals and soils. These processes have resulted in the release of toxic elements such as arsenic in groundwater aquifers, which is having a major impact on the health of millions of people in South and Southeast Asia. The interfaces between mineral surfaces and aqueous solutions are the locations of most chemical reactions that control the composition of the natural environment, including the composition of natural waters. The nuclear fuel cycle, from uranium mining to the disposition of high-level nuclear waste, is also intimately related to minerals. A fundamental understanding of these processes requires molecular-scale information about minerals, their bulk structures and properties such as solubility, their surfaces, and their interactions with aqueous solutions, atmospheric and soil gases, natural organic matter, and biological organisms. Gaining this understanding is further complicated by the presence of natural, incidental, and manufactured nanoparticles in the environment, which are

  14. 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

  15. Fuel behavior during a LOCA: LOFT experiments

    SciTech Connect

    Russell, M.L.

    1980-11-01

    The LOFT experiments have provided the following fuel behavior information which appears to be valuable for improving the safety of PWR operation and resolving PWR licensing issues: (1) A generic unassisted core cooling event occurs during large-break LOCAs that dominates the cooling of the core before ECC reflood commences and potentially eliminates the possibility of flow channel blockage from prepressurized fuel rod swelling. (2) The large-break LOCA decompression forces do not disturb the normal control rod gravity drop and may not structually damage the fuel assemblies. (3) Large-break LOCA core cooling may also be enhanced by spacer grid and core counter flow delay of liquid escape from the core boundaries and liquid fallback from the upper plenum into the core region. (4) Lower fuel rod prepressurization may be possible in PWR fuel rods which would reduce flow channel blockage complications during LOCA's. (5) Uniform fuel rod cladding temperature indications during the large break LOCA's do not confirm expectations for the fuel rod cladding temperature variations that would inhibit development of flow channel blockages by ballooning of prepressurized fuel rods.

  16. 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

  17. Failed MTR Fuel Element Detect in a Sipping Tests

    SciTech Connect

    Zeituni, C.A.; Terremoto, L.A.A.; da Silva, J.E.R.

    2004-10-06

    This work describes sipping tests performed on Material Testing Reactor (MTR) fuel elements of the IEA-R1 research reactor, in order to find out which one failed in the core during a routine operation. Radioactive iodine isotopes {sup 131}I and {sup 133}I, employed as failure monitors, were detected in samples corresponding to the failed fuel element. The specific activity of each sample, as well as the average leaking rate, were measured for {sup 137}Cs. The nuclear fuels U{sub 3}O{sub 8} - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of {sup 137}Cs.

  18. 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.

  19. 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.

  20. 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.

  1. 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)

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

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

    DOEpatents

    Gross, K.C.

    1988-01-21

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

  10. 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}.

  11. Method and apparatus for diagnosing breached fuel elements

    DOEpatents

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

    1988-01-01

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

  12. 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.

  13. 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.

  14. Characterization of Fuel Cell Vehicle Duty Cycle Elements

    SciTech Connect

    MAISH, ALEXANDER B.; NILAN, ERIC J.; BACA, PAUL M.

    2002-12-01

    This report covers research done as part of US Department of Energy contract DE-PS26-99FT14299 with the Fuel Cell Propulsion Institute on the fuel cell RATLER{trademark} vehicle, Lurch, as well as work done on the fuel cells designed for the vehicle. All work contained within this report was conducted at the Robotic Vehicle Range at Sandia National Laboratories in Albuquerque New Mexico. The research conducted includes characterization of the duty cycle of the robotic vehicle. This covers characterization of its various abilities such as hill climbing and descending, spin-turns, and driving on level ground. This was accomplished with the use of current sensors placed in the vehicle in conjunction with a Data Acquisition System (DAS), which was also created at Sandia Labs. Characterization of the two fuel cells was accomplished using various measuring instruments and techniques that will be discussed later in the report. A Statement of Work for this effort is included in Appendix A. This effort was able to complete characterization of vehicle duty cycle elements using battery power, but problems with the fuel cell control systems prevented completion of the characterization of the fuel cell operation on the benchtop and in the vehicle. Some data was obtained characterizing the fuel cell current-voltage performance and thermal rise rate by bypassing elements of the control system.

  15. 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)

  16. 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.

  17. PWR fuel behavior: lessons learned from LOFT. [PWR

    SciTech Connect

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior.

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

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

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

  19. 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

  20. 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)

  1. 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)

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    NASA Astrophysics Data System (ADS)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-09-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) or plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  8. Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect II: Effects of variations of the fuel particle diameters

    NASA Astrophysics Data System (ADS)

    Ding, Shurong; Wang, Qiming; Huo, Yongzhong

    2010-02-01

    In order to predict the irradiation mechanical behaviors of plate-type dispersion nuclear fuel elements, the total burnup is divided into two stages: the initial stage and the increasing stage. At the initial stage, the thermal effects induced by the high temperature differences between the operation temperatures and the room temperature are mainly considered; and at the increasing stage, the intense mechanical interactions between the fuel particles and the matrix due to the irradiation swelling of fuel particles are focused on. The large-deformation thermo-elasto-plasticity finite element analysis is performed to evaluate the effects of particle diameters on the in-pile mechanical behaviors of fuel elements. The research results indicate that: (1) the maximum Mises stresses and equivalent plastic strains at the matrix increase with the fuel particle diameters; the effects of particle diameters on the maximum first principal stresses vary with burnup, and the considered case with the largest particle diameter holds the maximum values all along; (2) at the cladding near the interface between the fuel meat and the cladding, the Mises stresses and the first principal stresses undergo major changes with increasing burnup, and different variations exist for different particle diameter cases; (3) the maximum Mises stresses at the fuel particles rise with the particle diameters.

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

    SciTech Connect

    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.

  10. The Behavior of Chemical Elements in Stars

    NASA Astrophysics Data System (ADS)

    Jaschek, Carlos; Jaschek, Mercedes

    1995-06-01

    Part I. Quantitative Description of Each of 80 Chemical Elements; Part II: 1. Behaviour of molecules in stars; 2. Groups of elements; 3. Chromospheres and coronas; Part III: 1. Terminology of spectral lines; 2. Selection of stars; 3. Line identification; 4. Equivalent widths; 5. Abundances; 6. Afterthoughts; Part IV: 1. Periodic Table; 2. Elements in alphabetical order of names; 3. Elements in alphabetical order of formula; 4. Elements ordered by atomic number; 5. Abundances of chemical elements; 6. Spectral type and surface gravity as a function of luminosity class. 7. Effective temperature; References; Index.

  11. The Behavior of Chemical Elements in Stars

    NASA Astrophysics Data System (ADS)

    Jaschek, Carlos; Jaschek, Mercedes

    2009-03-01

    Part I. Quantitative Description of Each of 80 Chemical Elements; Part II: 1. Behaviour of molecules in stars; 2. Groups of elements; 3. Chromospheres and coronas; Part III: 1. Terminology of spectral lines; 2. Selection of stars; 3. Line identification; 4. Equivalent widths; 5. Abundances; 6. Afterthoughts; Part IV: 1. Periodic Table; 2. Elements in alphabetical order of names; 3. Elements in alphabetical order of formula; 4. Elements ordered by atomic number; 5. Abundances of chemical elements; 6. Spectral type and surface gravity as a function of luminosity class. 7. Effective temperature; References; Index.

  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. 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)

  14. 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

  15. 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)

  16. Application of Thermochemical Modeling to Assessment/Evaluation of Nuclear Fuel Behavior

    SciTech Connect

    Besmann, Theodore M; McMurray, Jake W; Simunovic, Srdjan

    2016-01-01

    The combination of new fuel compositions and higher burn-ups envisioned for the future means that representing fuel properties will be much more important, and yet more complex. Behavior within the oxide fuel rods will be difficult to model owing to the high temperatures, and the large number of elements generated and their significant concentrations that are a result of fuels taken to high burn-up. This unprecedented complexity offers an enormous challenge to the thermochemical understanding of these systems and opportunities to advance solid solution models to describe these materials. This paper attempts to model and simulate that behavior using an oxide fuels thermochemical description to compute the equilibrium phase state and oxygen potential of LWR fuel under irradiation.

  17. Remote real time x-ray examination of fuel elements in a hot cell environment

    SciTech Connect

    Yapuncich, F.L.

    1993-03-01

    This report discusses the Remote Real Time X-ray System which will allow for detailed examination of fuel elements. This task will be accomplished in a highly radioactive hot cell environment. Two remote handling systems win be utilized at the examination station. One handling system will transfer the fuel element to and from the shielded x-ray system. A second handling system will allow for vertical and rotational inspection of the fuel elements. The process win include removing a single nuclear fuel element from a element fabrication magazine(EFM), positioning the fuel element within the shielding envelope of the x-ray system and transferring the fuel element from the station manipulator to the x-ray system manipulator, performing the x-ray inspection, and then transferring the fuel element to either the element storage magazine(ESM) or a reject bin.

  18. Remote real time x-ray examination of fuel elements in a hot cell environment

    SciTech Connect

    Yapuncich, F.L.

    1993-01-01

    This report discusses the Remote Real Time X-ray System which will allow for detailed examination of fuel elements. This task will be accomplished in a highly radioactive hot cell environment. Two remote handling systems win be utilized at the examination station. One handling system will transfer the fuel element to and from the shielded x-ray system. A second handling system will allow for vertical and rotational inspection of the fuel elements. The process win include removing a single nuclear fuel element from a element fabrication magazine(EFM), positioning the fuel element within the shielding envelope of the x-ray system and transferring the fuel element from the station manipulator to the x-ray system manipulator, performing the x-ray inspection, and then transferring the fuel element to either the element storage magazine(ESM) or a reject bin.

  19. Gamma-ray spectroscopy on irradiated MTR fuel elements

    NASA Astrophysics Data System (ADS)

    Terremoto, L. A. A.; Zeituni, C. A.; Perrotta, J. A.; da Silva, J. E. R.

    2000-08-01

    The availability of burnup data is an important requirement in any systematic approach to the enhancement of safety, economics and performance of a nuclear research reactor. This work presents the theory and experimental techniques applied to determine, by means of nondestructive gamma-ray spectroscopy, the burnup of Material Testing Reactor (MTR) fuel elements irradiated in the IEA-R1 research reactor. Burnup measurements, based on analysis of spectra that result from collimation and detection of gamma-rays emitted in the decay of radioactive fission products, were performed at the reactor pool area. The measuring system consists of a high-purity germanium (HPGe) detector together with suitable fast electronics and an on-line microcomputer data acquisition module. In order to achieve absolute burnup values, the detection set (collimator tube+HPGe detector) was previously calibrated in efficiency. The obtained burnup values are compared with ones provided by reactor physics calculations, for three kinds of MTR fuel elements with different cooling times, initial enrichment grades and total number of fuel plates. Both values show good agreement within the experimental error limits.

  20. Modeling Thermal and Stress Behavior of the Fuel-clad Interface in Monolithic Fuel Mini-plates

    SciTech Connect

    Gregory K. Miller; Pavel G. Medvedev; Douglas E. Burkes; Daniel M. Wachs

    2010-08-01

    As part of the Global Threat Reduction Initiative, a fuel development and qualification program is in process with the objective of qualifying very high density low enriched uranium fuel that will enable the conversion of high performance research reactors with operational requirements beyond those supported with currently available low enriched uranium fuels. The high density of the fuel is achieved by replacing the fuel meat with a single monolithic low enriched uranium-molybdenum fuel foil. Doing so creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. Furthermore, the monolithic fuel meat will dominate the structural properties of the fuel plate rather than the aluminum matrix, which is characteristic of dispersion fuel types. Understanding the integrity and behavior of the fuel-clad interface during irradiation is of great importance for qualification of the new fuel, but can be somewhat challenging to determine with a single technique. Efforts aimed at addressing this problem are underway within the fuel development and qualification program, comprised of modeling, as-fabricated plate characterization, and post-irradiation examination. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure, using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initial analysis shows that the stress normal to the fuel-clad interface dominates during irradiation, and that the presence of small, rounded delaminations at the interface is not of great concern. However, larger and/or fuel-clad delaminations with sharp corners can create areas of concern, as maximum principal cladding stress, strain, displacement, and peak fuel temperature are all significantly increased. Furthermore, stresses resulting from temperature gradients that cause the plate to bow or buckle in

  1. Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect I: Effects of variations of the fuel particle volume fractions

    NASA Astrophysics Data System (ADS)

    Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

    2010-05-01

    A new method of modeling the in-pile mechanical behaviors of dispersion nuclear fuel elements is proposed. Considering the irradiation swelling together with the thermal effect, numerical simulations of the in-pile mechanical behaviors are performed with the developed finite element models for different fuel particle volume fractions of the fuel meat. The effects of the particle volume fractions on the mechanical performances of the fuel element are studied. The research results indicate that: (1) the maximum Mises stresses and equivalent plastic strains at the matrix increase with the particle volume fractions at each burnup; the locations of the maximum first principal stresses shift with increasing burnup; at low burnups, the maximum first principal stresses increase with the particle volume fractions; while at high burnups, the 20% volume fraction case holds the lowest value; (2) at the cladding, the maximum equivalent plastic strains and the tensile principal stresses increase with the particle volume fractions; while the maximum Mises stresses do not follow this order at high burnups; (3) the maximum Mises stresses at the fuel particles increase with the particle volume fractions, and the particles will engender plastic strains until the particle volume fraction reaches high enough.

  2. 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

  3. Combustion behavior of solid fuel ramjets

    NASA Technical Reports Server (NTRS)

    Netzer, D. W.; Binn, B. A.; Scott, W. E.; Metochianakis, M.

    1980-01-01

    Nonreacting flowfield characteristics and fundamental fuel properties are considered with respect to their use in estimating the obtainable combustion efficiency for fuels and/or combustor geometries. It is shown that near wall turbulence intensity in nonreacting flow appears to correlate reasonably well with the fuel regression pattern in identical geometries. The HTPB based fuels exhibit solid phase exothermic reactions in contrast to purely endothermic reactions for plexiglas. It is further shown that combustion pressure oscillations appear to be related to physically induced disturbances to the fluctuating shear layers at the fuel grain and aft mixing chamber inlets.

  4. 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

  5. Neutronics and fuel behavior of AIROX-processed fuel recycled into light water reactors

    SciTech Connect

    Allison, C.M.; Jahshan, S.N.; Wade, N.L.

    1993-08-01

    An evaluation of the Atomics International Reduction Oxidation (AIROX) process has begun to determine if the process could be used to recycle spent fuel to minimize high-level waste from commercial power reactors. This paper includes an evaluation of core neutronics to establish enrichment levels and expected in-reactor performance: a review of existing fuel behavior research to determine its applicability to AIROX-recycled fuels; and an evaluation of potential licensing issues unique to these fuels.

  6. Writing Comprehensive Behavioral Consultation Reports: Critical Elements

    ERIC Educational Resources Information Center

    Brinkman, Tara M.; Segool, Natasha K.; Pham, Andy V.; Carlson, John S.

    2007-01-01

    The accountability movement in psychology has resulted in practitioners increasingly using evidence-based interventions and treatment modalities to treat client problems. Behavioral consultation is one framework that practitioners can utilize in providing empirically supported services. In order to demonstrate the use of effective, evidence-based…

  7. Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

    NASA Astrophysics Data System (ADS)

    Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

    2009-08-01

    Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

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

    SciTech Connect

    Long, Jr. E.L.

    2001-10-25

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

  9. 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.

  10. 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.

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

    PubMed

    Saqib, Naeem; Bäckström, Mattias

    2015-10-01

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

  12. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    NASA Astrophysics Data System (ADS)

    Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

    2011-06-01

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  13. 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.

  14. Fuel-cladding interaction layers in irradiated U-ZR and U-PU-ZR fuel elements.

    SciTech Connect

    Keiser, D. D.

    2006-01-23

    Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr and U-Pu-Zr alloy fuel elements irradiated in the Experimental Breeder Reactor-II (EBR-II). The electrometallurgical treatment process extracts usable uranium from irradiated fuel elements and places residual fission products, actinides, process Zr, and cladding hulls (small segments of tubing) into two waste forms--a ceramic and a metal alloy. The metal waste form will contain the cladding hulls, Zr, and noble metal fission products, and it will be disposed of in a geologic repository. As a result, the expected composition of the waste form will need to be well understood. This report deals with the condition of the cladding, which will make up a large fraction of the metal waste form, after irradiation in EBR-II and before insertion into the electrorefiner. Specifically, it looks at layers that can be found on the inner surface of the cladding due to in-reactor interactions between the alloy fuel and the stainless steel cladding that occurs after the fuel has swelled and contacted the cladding. Many detailed examinations of fuel elements irradiated in EBR-II have been completed and are discussed in the context of interaction layer formation in irradiated cladding. The composition and thickness of the developed interaction layers are identified, along with the irradiation conditions, cladding type, and axial location on fuel elements where the thickest interaction layers can be expected to develop. It has been found that the largest interaction zones are observed at combined high power and high temperature regions of fuel elements and for fuel elements with U-Pu-Zr alloy fuel and D9 stainless steel cladding. The most prevalent, non-cladding constituent observed in the developed interaction layers are the lanthanide fission products.

  15. 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.

  16. WREM--TOODEE2--MOD3. 2d Time-Dependent Fuel Element Study

    SciTech Connect

    Lauben, G.N.

    1992-03-05

    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).

  17. 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.

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

    NASA Astrophysics Data System (ADS)

    Bockstahler, Klaus; Funke, Helmut; Lucas, Joachim

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

  19. Composition and Behavior of Fuel Ethanol

    EPA Science Inventory

    Ethanol usage in the United States has increased due in part to the elimination of methyl tert-butyl ether from the fuel supply and to the mandates of Congress. Two samples, one each from a wet mill and a dry mill ethanol plant, were obtained before denaturing. Each of these ...

  20. 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.

  1. Structural Behavior of Monolithic Fuel Plates During Hot Isostatic Pressing and Annealing

    SciTech Connect

    Pavel G. Medvedev; Hakan Ozaltun

    2010-03-01

    This paper presents results of the stress analysis in the monolithic fuel plates during thermal transients performed using COMSOL finite element analysis software. Large difference in the thermal expansion between the U-Mo foil and Al cladding is the main load origin during heating and cooling of the fuel plates. In addition, the mechanical behavior of the plate is affected by the difference in yield points between the foil and the cladding. This is manifested by the plastic deformation and permanent strains in the cladding, and elastic deformation of the foil. The results show existence of the critical temperature points at which the stresses change from compressive to tensile. The paper highlights principal differences in mechanical behavior between monolithic and dispersion fuel plates, underlines the need for mechanical property data, especially for the U-Mo alloys, and discusses the methodology for mechanical analysis of the monolithic plates.

  2. 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).

  3. Behavioral Health and Performance Element: Tools and Technologies

    NASA Technical Reports Server (NTRS)

    Leveton, Lauren B.

    2009-01-01

    This slide presentation reviews the research into the Behavioral Health and Performance (BHP) of the Human Research Program. The program element goal is to identify, characterize and prevent or reduce behavioral health and performance risks associated with space travel, exploration, and return to terrestrial life. To accomplish this goal the program focuses on applied research that is designed to yield deliverables that reduce risk. There are several different elements that are of particular interest: Behavioral Medicine, Sleep, and team composition, and team work. In order to assure success for NASA missions the Human Research Program develops and validate the standards for each of the areas of interest. There is discussion of the impact on BHP while astronauts are on Long Duration Missions. The effort in this research is to create tools to meet the BHP concerns, these prospective tools are reviewed.

  4. Criticality safety evaluation for pathfinder fuel elements in model No. RA-3 shipping containers

    SciTech Connect

    Jones, R.R.

    1986-11-01

    Pennsylvania State University presently processes approximately 415 Pathfinder fuel elements which will require shipment from their nuclear facility. Criticality safety calculations have been performed with the Monte Carlo code, KENO-IV, and 16-group Hansen-Roach cross sections for shipment of these fuel elements in Model No. RA-3 shipping containers. Except for a slightly higher U-235 enrichment in the UO/sub 2/ rods of the Pathfinder fuel elements, the parameters for the proposed shipment are within those limits currently approved in Certificate of Compliance No. 4986, Revision No. 17, for shipment of UO/sub 2/ fuel rods in the Model RA-3 shipping containers. The analysis in this report verifies an adequate margin of criticality safety for the Pathfinder fuel elements in Model RA-3 containers for a Fissile Class 1 shipment.

  5. 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.

  6. 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.

  7. 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.

  8. Deposition behavior of UO2 and noble-metal elements in oxide-electrowinning reprocessing

    NASA Astrophysics Data System (ADS)

    Kosugi, K.; Fukushima, M.; Myochin, M.; Mizuguchi, K.; Oomori, T.

    2005-02-01

    As a candidate process for future reprocessing technology of nuclear spent fuel, oxide-electrowinning method has been studied. In this method, the uranium is collected on the cathode in the form of UO2 by electrolysis in the molten chloride. Thereby, the noble metal (NM) elements accompany the uranium deposition, because of very close redox potential between NM elements and UO2. To clarify the electrolysis behavior of the uranium and NM elements in the low-current-density electrolysis, the laboratory scale experiments were performed under various conditions of cathode current density and solutes concentration in the chloride melt, and the separation efficiency and the morphology of the deposition were investigated. It was found that the separation of Pd from uranium was more difficult than that of Rh. The presence of U4+ greatly influenced current efficiency of the electrolysis process.

  9. Single-element coaxial injector for rocket fuel

    NASA Technical Reports Server (NTRS)

    Larson, L. L.

    1969-01-01

    Improved injector for oxygen difluoride and diborane has better mixing characteristics and is able to project fuel onto the wall of the combustion chamber for better cooling. It produces an essentially conical, diverging, continuous sheet of propellant mixture formed by similarly shaped and continuously impinging sheets of fuel and oxidant.

  10. EVALUATION OF U10MO FUEL PLATE IRRADIATION BEHAVIOR VIA NUMERICAL AND EXPERIMENTAL BENCHMARKING

    SciTech Connect

    Samuel J. Miller; Hakan Ozaltun

    2012-11-01

    This article analyzes dimensional changes due to irradiation of monolithic plate-type nuclear fuel and compares results with finite element analysis of the plates during fabrication and irradiation. Monolithic fuel plates tested in the Advanced Test Reactor (ATR) at Idaho National Lab (INL) are being used to benchmark proposed fuel performance for several high power research reactors. Post-irradiation metallographic images of plates sectioned at the midpoint were analyzed to determine dimensional changes of the fuel and the cladding response. A constitutive model of the fabrication process and irradiation behavior of the tested plates was developed using the general purpose commercial finite element analysis package, Abaqus. Using calculated burn-up profiles of irradiated plates to model the power distribution and including irradiation behaviors such as swelling and irradiation enhanced creep, model simulations allow analysis of plate parameters that are either impossible or infeasible in an experimental setting. The development and progression of fabrication induced stress concentrations at the plate edges was of primary interest, as these locations have a unique stress profile during irradiation. Additionally, comparison between 2D and 3D models was performed to optimize analysis methodology. In particular, the ability of 2D and 3D models account for out of plane stresses which result in 3-dimensional creep behavior that is a product of these components. Results show that assumptions made in 2D models for the out-of-plane stresses and strains cannot capture the 3-dimensional physics accurately and thus 2D approximations are not computationally accurate. Stress-strain fields are dependent on plate geometry and irradiation conditions, thus, if stress based criteria is used to predict plate behavior (as opposed to material impurities, fine micro-structural defects, or sharp power gradients), unique 3D finite element formulation for each plate is required.

  11. Elemental Solubility Tendency for the Phases of Uranium by Classical Models Used to Predict Alloy Behavior

    SciTech Connect

    Van Blackwood; Travis Koenig; Saleem Drera; Brajenda Mishra; Davis Olson; Doug Porter; Robert Mariani

    2012-03-01

    Traditional alloy theory models, specifically Darken-Gurry and Miedema’s analyses, that characterize solutes in solid solvents relative to physical properties of the elements have been used to assist in predicting alloy behavior. These models will be applied relative to the three solid phases of uranium: alpha (orthorhombic), beta (tetragonal), and gamma (bcc). These phases have different solubilities for specific alloy additions as a function of temperature. The Darken-Gurry and Miedema models, with modifications based on concepts of Waber, Gschneider, and Brewer will be used to predict the behavior of four types of solutes: 1) Transition metals that are used for various purposes associated with the containment as alloy additions in the uranium fuel 2) Transuranic elements in the uranium 3) Rare earth fission products (lanthanides) 4) Transition metals and other fission products Using these solute map criteria, elemental behavior will be predicted as highly soluble, marginally soluble, or immiscible (compound formers) and will be used to compare solute effects during uranium phase transformations. The overlapping of these solute maps are convenient first approximation tools for predicting alloy behavior.

  12. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    NASA Astrophysics Data System (ADS)

    Makmal, T.; Aviv, O.; Gilad, E.

    2016-10-01

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.

  13. A MULTIDIMENSIONAL AND MULTIPHYSICS APPROACH TO NUCLEAR FUEL BEHAVIOR SIMULATION

    SciTech Connect

    R. L. Williamson; J. D. Hales; S. R. Novascone; M. R. Tonks; D. R. Gaston; C. J. Permann; D. Andrs; R. C. Martineau

    2012-04-01

    Important aspects of fuel rod behavior, for example pellet-clad mechanical interaction (PCMI), fuel fracture, oxide formation, non-axisymmetric cooling, and response to fuel manufacturing defects, are inherently multidimensional in addition to being complicated multiphysics problems. Many current modeling tools are strictly 2D axisymmetric or even 1.5D. This paper outlines the capabilities of a new fuel modeling tool able to analyze either 2D axisymmetric or fully 3D models. These capabilities include temperature-dependent thermal conductivity of fuel; swelling and densification; fuel creep; pellet fracture; fission gas release; cladding creep; irradiation growth; and gap mechanics (contact and gap heat transfer). The need for multiphysics, multidimensional modeling is then demonstrated through a discussion of results for a set of example problems. The first, a 10-pellet rodlet, demonstrates the viability of the solution method employed. This example highlights the effect of our smeared cracking model and also shows the multidimensional nature of discrete fuel pellet modeling. The second example relies on our the multidimensional, multiphysics approach to analyze a missing pellet surface problem. As a final example, we show a lower-length-scale simulation coupled to a continuum-scale simulation.

  14. 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

  15. Dialectical behavior therapy: current indications and unique elements.

    PubMed

    Chapman, Alexander L

    2006-09-01

    Dialectical behavior therapy (DBT) is a comprehensive, evidence-based treatment for borderline personality disorder (BPD). The patient populations for which DBT has the most empirical support include parasuicidal women with borderline personality disorder (BPD), but there have been promising findings for patients with BPD and substance use disorders (SUDs), persons who meet criteria for binge-eating disorder, and depressed elderly patients. Although DBT has many similarities with other cognitive-behavioral approaches, several critical and unique elements must be in place for the treatment to constitute DBT. Some of these elements include (a) serving the five functions of treatment, (b) the biosocial theory and focusing on emotions in treatment, (c) a consistent dialectical philosophy, and (d) mindfulness and acceptance-oriented interventions.

  16. Spent fuel behavior under abnormal thermal transients during dry storage

    SciTech Connect

    Stahl, D.; Landow, M.P.; Burian, R.J.; Pasupathi, V.

    1986-01-01

    This study was performed to determine the effects of abnormally high temperatures on spent fuel behavior. Prior to testing, calculations using the CIRFI3 code were used to determine the steady-state fuel and cask component temperatures. The TRUMP code was used to determine transient heating rates under postulated abnormal events during which convection cooling of the cask surfaces was obstructed by a debris bed covering the cask. The peak rate of temperature rise during the first 6 h was calculated to be about 15/sup 0/C/h, followed by a rate of about 1/sup 0/C/h. A Turkey Point spent fuel rod segment was heated to approx. 800/sup 0/C. The segment deformed uniformly with an average strain of 17% at failure and a local strain of 60%. Pretest characterization of the spent fuel consisted of visual examination, profilometry, eddy-current examination, gamma scanning, fission gas collection, void volume measurement, fission gas analysis, hydrogen analysis of the cladding, burnup analysis, cladding metallography, and fuel ceramography. Post-test characterization showed that the failure was a pinhole cladding breach. The results of the tests showed that spent fuel temperatures in excess of 700/sup 0/C are required to produce a cladding breach in fuel rods pressurized to 500 psing (3.45 MPa) under postulated abnormal thermal transient cask conditions. The pinhole cladding breach that developed would be too small to compromise the confinement of spent fuel particles during an abnormal event or after normal cooling conditions are restored. This behavior is similar to that found in other slow ramp tests with irradiated and nonirradiated rod sections and nonirradiated whole rods under conditions that bracketed postulated abnormal heating rates. This similarity is attributed to annealing of the irradiation-strengthened Zircaloy cladding during heating. In both cases, the failure was a benign, ductile pinhole rupture.

  17. Biogeochemistry of Hot Spring Biofilms: Major and Trace Element Behavior

    NASA Astrophysics Data System (ADS)

    Havig, J. R.; Prapaipong, P.; Zolotova, N.; Moore, G.; Shock, E. L.

    2008-12-01

    Hot spring biofilms are of obvious biological origin, but of surprising composition. Organic carbon makes up a minor percentage of the total mass of chemotrophic and phototrophic biofilms. We have found that the majority of biofilm mass is inorganic material, largely silica, with measurable quantities of dozens of other elements, and that the distribution of major elements mimics that of surrounding rock and soil far more closely than the hot spring fluids. Comparisons of biofilms with the compositions of their geochemical surroundings help identify trace elements that are anomalously enriched or depleted. These anomalies provide insight into the processes of active or passive elemental accumulation by biofilms, which could be used to understand microbial processes of element uptake or to identify evidence for life in hydrothermal deposits in the rock record. Five separate hydrothermal systems in Yellowstone National Park were incorporated into this study: 'Bison Pool' and its outflow (siliceous-sinter depositing, temp. = 93.2 to 56.2 C, pH = 7.4 to 8.3), Flatcone Geyser and its outflow (siliceous-sinter depositing, temp. = 94.3 to 44.3 C, pH = 7.9 to 8.8, Boulder Spring and its outflow (siliceous-sinter depositing, temp. = 92.1 to 64.9 C, pH = 8.2 to 8.7), Octopus Spring and its outflow (siliceous-sinter depositing, temp. = 91.4 to 62.8 C, pH = 7.7 to 8.2), and two unnamed locations in the Obsidian Pool area we have dubbed 'Green Cheese' (temp. = 64.5 to 54.9 C, pH = 5.9 to 6.2) and 'Happy Harfer Pool' (temp. = 59.9 to 48.3 C, pH = 5.5 to 6.3). Analysis of water, biofilm, and contextual samples collected from and around these hot springs offer intriguing patterns of elemental behavior, both similar and dissimilar, among the varying systems. Examples of these patterns include elements that behave the same across all hot spring systems (B, C, Ni, Cu, Ge, Sb, and W), elements with behavior that was consistent throughout most (four of five) of the hot spring systems

  18. A computational study of nodal-based tetrahedral element behavior.

    SciTech Connect

    Gullerud, Arne S.

    2010-09-01

    This report explores the behavior of nodal-based tetrahedral elements on six sample problems, and compares their solution to that of a corresponding hexahedral mesh. The problems demonstrate that while certain aspects of the solution field for the nodal-based tetrahedrons provide good quality results, the pressure field tends to be of poor quality. Results appear to be strongly affected by the connectivity of the tetrahedral elements. Simulations that rely on the pressure field, such as those which use material models that are dependent on the pressure (e.g. equation-of-state models), can generate erroneous results. Remeshing can also be strongly affected by these issues. The nodal-based test elements as they currently stand need to be used with caution to ensure that their numerical deficiencies do not adversely affect critical values of interest.

  19. 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.

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

    PubMed

    Saqib, Naeem; Bäckström, Mattias

    2015-10-01

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

  1. 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)

  2. METHOD OF MAKING A COMPARTMENTED FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    McGeary, R.K.; Frisch, E.

    1963-05-14

    A method of making a compartmented fuel element is presented. Fuel pellets arid spacing disks are inserted into a cladding tube; plugs are inserted at each end and the tube is then stretched lengthwise so that its walls grip the edges of the spacing disks, thereby forming compartments. (AEC)

  3. Trace element partitioning behavior of coal gangue-fired CFB plant: experimental and equilibrium calculation.

    PubMed

    Zhang, Yingyi; Nakano, Jinichiro; Liu, Lili; Wang, Xidong; Zhang, Zuotai

    2015-10-01

    Energy recovery is a promising method for coal gangue utilization, during which the prevention of secondary pollution, especially toxic metal emission, is a significant issue in the development of coal gangue utilization. In the present study, investigation into trace element partitioning behavior from a coal gangue-fired power plant in Shanxi province, China, has been conducted. Besides the experimental analysis, thermodynamic equilibrium calculation was also conducted to help the further understanding on the effect of different parameters. Results showed that Hg, As, Be, and Cd were highly volatile elements in the combustion of coal gangue, which were notably enriched in fly ash and may be emitted into the environment via the gas phase. Cr and Mn were mostly non-volatile and were enriched in the bottom ash. Pb, Co, Zn, Cu, and Ni were semi-volatile elements and were enriched in the fly ash to varying degrees. Equilibrium calculations show that the air/fuel ratio and the presence of Cl highly affect the element volatility. The presence of mineral phases, such as aluminosilicates, depresses the volatility of elements by chemical immobilization and competition in Cl. The coal gangue, fly ash, and bottom ash all passed the toxicity characteristic leaching procedure (TCLP), and their alkalinity buffers the acidity of the solution and contributes to the low solubility of the trace elements. PMID:26006077

  4. Trace element partitioning behavior of coal gangue-fired CFB plant: experimental and equilibrium calculation.

    PubMed

    Zhang, Yingyi; Nakano, Jinichiro; Liu, Lili; Wang, Xidong; Zhang, Zuotai

    2015-10-01

    Energy recovery is a promising method for coal gangue utilization, during which the prevention of secondary pollution, especially toxic metal emission, is a significant issue in the development of coal gangue utilization. In the present study, investigation into trace element partitioning behavior from a coal gangue-fired power plant in Shanxi province, China, has been conducted. Besides the experimental analysis, thermodynamic equilibrium calculation was also conducted to help the further understanding on the effect of different parameters. Results showed that Hg, As, Be, and Cd were highly volatile elements in the combustion of coal gangue, which were notably enriched in fly ash and may be emitted into the environment via the gas phase. Cr and Mn were mostly non-volatile and were enriched in the bottom ash. Pb, Co, Zn, Cu, and Ni were semi-volatile elements and were enriched in the fly ash to varying degrees. Equilibrium calculations show that the air/fuel ratio and the presence of Cl highly affect the element volatility. The presence of mineral phases, such as aluminosilicates, depresses the volatility of elements by chemical immobilization and competition in Cl. The coal gangue, fly ash, and bottom ash all passed the toxicity characteristic leaching procedure (TCLP), and their alkalinity buffers the acidity of the solution and contributes to the low solubility of the trace elements.

  5. Quantifying Square Membrane Wrinkle Behavior Using MITC Shell Elements

    NASA Technical Reports Server (NTRS)

    Jacobson, Mindy B.; Iwasa, Takashi; Natori, M. C.

    2004-01-01

    For future membrane based structures, quantified predictions of membrane wrinkling behavior in terms of amplitude, angle and wavelength are needed to optimize the efficiency and integrity of such structures, as well as their associated control systems. For numerical analyses performed in the past, limitations on the accuracy of membrane distortion simulations have often been related to the assumptions made while using finite elements. Specifically, this work demonstrates that critical assumptions include: effects of gravity. supposed initial or boundary conditions, and the type of element used to model the membrane. In this work, a 0.2 square meter membrane is treated as a structural material with non-negligible bending stiffness. Mixed Interpolation of Tensorial Components (MTTC) shell elements are used to simulate wrinkling behavior due to a constant applied in-plane shear load. Membrane thickness, gravity effects, and initial imperfections with respect to flatness were varied in numerous nonlinear analysis cases. Significant findings include notable variations in wrinkle modes for thickness in the range of 50 microns to 1000 microns, which also depend on the presence of an applied gravity field. However, it is revealed that relationships between overall strain energy density for cases with differing initial conditions are independent of assumed initial con&tions. In addition, analysis results indicate that the relationship between amplitude scale (W/t) and structural scale (L/t) is linear in the presence of a gravity field.

  6. 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.

  7. 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.

  8. 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.

  9. Criticality safety evaluation for pathfinder fuel elements in model No. RA-3 shipping containers

    SciTech Connect

    Jones, R.R.

    1990-02-01

    Pennsylvania State University presently possesses approximately 415 Pathfinder fuel elements which require shipment from their nuclear facility. It is planned to use Model No. RA-3 shipping containers for shipment of these elements. Certificate of Compliance No. 4986, Revision No. 22, Docket No. 71-4986 authorizes the use of these containers for Fissile Class 1 shipments of UO{sub 2} fuel assemblies and UO{sub 2} fuel rods with parameters similar to the Pathfinder fuel elements. Criticality safety calculations have been performed with the Monte Carlo code, KENO-V.a and 16-group Hansen-Roach cross sections for shipment of Pathfinder fuel elements in Model No. RA-3 shipping containers. The analysis is described and the results are given in this report. This analysis demonstrates that the RA-3 container with Pathfinder fuel elements complies with the requirements of 10 CFR 71.59 for Fissile Class 2 shipping containers with six as the allowable number of containers in a single shipment. 5 refs., 5 tabs., 6 figs.

  10. Criticality safety evaluation for Pathfinder fuel elements in Model No. RA-3 shipping containers: Revision 1

    SciTech Connect

    Jones, R.R.

    1989-01-01

    Pennsylvania State University presently possesses approximately 415 Pathfinder fuel elements which require shipment from their nuclear facility. It is planned to use Model No. RA-3 shipping containers for shipment of these elements. Certificate of Compliance No. 4986, Revision No. 18, Docket No. 71-4986 authorizes the use of these containers for Fissile Class I shipments of UO/sub 2/ fuel assemblies and UO/sub 2/ fuel rods with parameters similar to the Pathfinder fuel elements. Criticality safety calculations have been performed with the Monte Carlo code, KENO-V.a and 16-group Hansen-Roach cross sections for shipment of Pathfinder fuel elements in Model No. RA-3 shipping containers. The analysis is described and the results are given in this report. This analysis demonstrates that the RA-3 container with Pathfinder fuel elements complies with the requirements of 10 CFR 71.59 for Fissile Class II shipping containers with six as the allowable number of containers in a single shipment. 5 refs., 6 figs., 4 tabs.

  11. Prediction of the micro-thermo-mechanical behaviors in dispersion nuclear fuel plates with heterogeneous particle distributions

    NASA Astrophysics Data System (ADS)

    Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong; Zhang, Lin; Li, Yuanming

    2011-11-01

    Dispersion nuclear fuel elements have promising prospects to be used in advanced nuclear reactors and disposal of nuclear wastes. They consist of fuel meat and cladding, and the fuel meat is a kind of composite fuel in which the fuel particles are embedded in the non-fissile matrix. Prediction of the micro-thermo-mechanical behaviors in dispersion nuclear plates is of importance to their irradiation safety and optimal design. In this study, the heterogeneity of the fuel particles along the thickness direction in the fuel meat is considered. The 3D finite element models have been developed respectively for two cases: (1) variation of fuel particle-particle (PP) distances for the particles near the mid-plane of the fuel meat; (2) variation of the particle-cladding (PC) distances for the fuel particles near the interface between the fuel meat and the cladding. The respective finite strain constitutive relations are developed for the fuel particle, metal matrix and cladding. The developed virtual temperature method is used to simulate irradiation swelling of the fuel particles and irradiation growth of the metal cladding. Effects of the heterogeneous distributions of the fuel particles on the micro temperature fields and the micro stress-strain fields are investigated. The obtained results indicate that: (1) as a whole, the maximum Mises stress, equivalent plastic strain and first principal stress at the matrix between the two closest particles increase with decreasing the particle-particle (PP) distance; existence of large first principal stresses there may be the main factor that induces the matrix failure; (2) variation of the particle-cladding (PC) distance has remarkable effects on the interfacial normal stress and shear stress at the interface between the fuel meat and the cladding; the first principal stress at the cladding near the interface increases dramatically when the fuel particle is closer and closer to the cladding. Thus, the proper distance between the

  12. 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.

  13. Behavioral economic analysis of demand for fuel in North America.

    PubMed

    Reed, Derek D; Partington, Scott W; Kaplan, Brent A; Roma, Peter G; Hursh, Steven R

    2013-01-01

    Emerging research clearly indicates that human behavior is contributing to climate change, notably, the use of fossil fuels as a form of energy for everyday behaviors. This dependence on oil in North America has led to assertions that the current level of demand is the social equivalent to an "addiction." The purpose of this study was to apply behavioral economic demand curves-a broadly applicable method of evaluating relative reinforcer efficacy in behavioral models of addiction-to North American oil consumption to examine whether such claims of oil addiction are warranted. Toward this end, we examined government data from the United States and Canada on per capita energy consumption for transportation and oil prices between 1995 and 2008. Our findings indicate that consumption either persisted or simultaneously increased despite sharp increases in oil price per barrel over the past decade.

  14. Burn-up and Operation Time of Fuel Elements Produced in IPEN

    NASA Astrophysics Data System (ADS)

    Tondin, Julio Benedito Marin; Filho, Tufic Madi

    2011-08-01

    The aim of this paper is to present the developed work along the operational and reliability tests of fuel elements produced in the Institute of Energetic and Nuclear Research, IPEN-CNEN/SP, from the 1980's. The study analyzed the U-235 burn evolution and the element remain in the research reactor IEA-R1. The fuel elements are of the type MTR (Material Testing Reactor), the standard with 18 plates and a 12-plate control, with a nominal mean enrichment of 20%.

  15. Diagnosis of PEM fuel cell stack dynamic behaviors

    NASA Astrophysics Data System (ADS)

    Chen, Jixin; Zhou, Biao

    In this study, the steady-state performance and dynamic behavior of a commercial 10-cell Proton Exchange Membrane (PEM) fuel cell stack was experimentally investigated using a self-developed PEM fuel cell test stand. The start-up characteristics of the stack to different current loads and dynamic responses after current step-up to an elevated load were investigated. The stack voltage was observed to experience oscillation at air excess coefficient of 2 due to the flooding/recovery cycle of part of the cells. In order to correlate the stack voltage with the pressure drop across the cathode/anode, fast Fourier transform was performed. Dominant frequency of pressure drop signal was obtained to indicate the water behavior in cathode/anode, thereby predicting the stack voltage change. Such relationship between frequency of pressure drop and stack voltage was found and summarized. This provides an innovative approach to utilize frequency of pressure drop signal as a diagnostic tool for PEM fuel cell stack dynamic behaviors.

  16. 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.

  17. Integrating Health Behavior Theory and Design Elements in Serious Games

    PubMed Central

    Fleming, Theresa; Lucassen, Mathijs FG; Bridgman, Heather; Stasiak, Karolina; Shepherd, Matthew; Orpin, Peter

    2015-01-01

    Background Internet interventions for improving health and well-being have the potential to reach many people and fill gaps in service provision. Serious gaming interfaces provide opportunities to optimize user adherence and impact. Health interventions based in theory and evidence and tailored to psychological constructs have been found to be more effective to promote behavior change. Defining the design elements which engage users and help them to meet their goals can contribute to better informed serious games. Objective To elucidate design elements important in SPARX, a serious game for adolescents with depression, from a user-centered perspective. Methods We proposed a model based on an established theory of health behavior change and practical features of serious game design to organize ideas and rationale. We analyzed data from 5 studies comprising a total of 22 focus groups and 66 semistructured interviews conducted with youth and families in New Zealand and Australia who had viewed or used SPARX. User perceptions of the game were applied to this framework. Results A coherent framework was established using the three constructs of self-determination theory (SDT), autonomy, competence, and relatedness, to organize user perceptions and design elements within four areas important in design: computer game, accessibility, working alliance, and learning in immersion. User perceptions mapped well to the framework, which may assist developers in understanding the context of user needs. By mapping these elements against the constructs of SDT, we were able to propose a sound theoretical base for the model. Conclusions This study’s method allowed for the articulation of design elements in a serious game from a user-centered perspective within a coherent overarching framework. The framework can be used to deliberately incorporate serious game design elements that support a user’s sense of autonomy, competence, and relatedness, key constructs which have been found

  18. 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.

  19. Magnesium transport extraction of transuranium elements from LWR fuel

    SciTech Connect

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

    1992-09-15

    This patent describes a process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuel containing rare earth and noble metal fission products as well as fission products of alkali metals, alkaline earth metals and iodine. It comprises reducing the oxide fuel with Ca metal in the presence of Ca halide; separating the Ca halide with the CaO and the fission products contained therein from the U-Fe alloy and the metal values dissolved therein and electrolytically contacting the calcium salts with a carbon electrode; contacting the liquid U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel with liquid Mg metal, thereafter separating the liquid Mg and the metals dissolved therein from the U-Fe alloy and the metal dissolved therein, distilling the Mg from the transuranium actinide and rare earth metals, recontacting the U-Fe alloy with liquid Mg metal a sufficient number of times until not less than about 99% by weight of the transuranium actinide values have been removed from the U-Fe alloy.

  20. Uranium chloride extraction of transuranium elements from LWR fuel

    SciTech Connect

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

    1991-12-31

    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{degrees}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. Magnesium transport extraction of transuranium elements from LWR fuel

    SciTech Connect

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

    1991-12-31

    This report discusses 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{degrees}C to about 850{degrees}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.

  2. Uranium chloride extraction of transuranium elements from LWR fuel

    DOEpatents

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

    1992-01-01

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

  3. Magnesium transport extraction of transuranium elements from LWR fuel

    DOEpatents

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

    1992-01-01

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

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

    DOEpatents

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

    1981-04-28

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

  5. 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.

  6. 77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-22

    ... COMMISSION Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test... draft regulatory guide (DG), DG-2005, ``Quality Verification for Plate-Type Uranium-Aluminum Fuel... quality assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used...

  7. The Behavior of Orbital Element of 1566 Icarus Asteroids

    NASA Astrophysics Data System (ADS)

    Soegiartini, Endang; Radiman, Iratius; Fauzi, Umar; Siregar, Suryadi

    1566 Icarus is an asteroid with special orbital elements; high eccentricity (e = 0.8269), high incli-nation (i = 22o .8368), small semimajor-axis (a = 1.0778 AU), argument of perihelion (31o .3393), longitude of ascending node (88o .0474), and mean anomaly (M = 85o .8306) at epoch 2455200.5 JD or January 4, 2010. In this paper, we would like to trace the orbital evolution of 1566 Icarus for 200,000 years, from 100,000 BC until 100,000 AD. The gravitational influence of the eight planets was included in the integrations, which were all carried out using the hybrid inte-grator within the Mercury 6 (Chambers, 1999) software package. From this, we try to predict the behavior of its orbital element by making 243 (35 ) clones of 1566 Icarus. The clones are made by repeated permutation of nominal, nominal+1σ, nominal-1σ with the orbital elements: semi-major axis, eccentricity, orbital inclination, longitude of ascending node, and argument of perihelion. From this cloning, we see that the orbital evolution of 1566 Icarus will be stable for 200,000 years, but with small variation in semi-major axis (a).

  8. Predicting the behavior of microfluidic circuits made from discrete elements

    PubMed Central

    Bhargava, Krisna C.; Thompson, Bryant; Iqbal, Danish; Malmstadt, Noah

    2015-01-01

    Microfluidic devices can be used to execute a variety of continuous flow analytical and synthetic chemistry protocols with a great degree of precision. The growing availability of additive manufacturing has enabled the design of microfluidic devices with new functionality and complexity. However, these devices are prone to larger manufacturing variation than is typical of those made with micromachining or soft lithography. In this report, we demonstrate a design-for-manufacturing workflow that addresses performance variation at the microfluidic element and circuit level, in context of mass-manufacturing and additive manufacturing. Our approach relies on discrete microfluidic elements that are characterized by their terminal hydraulic resistance and associated tolerance. Network analysis is employed to construct simple analytical design rules for model microfluidic circuits. Monte Carlo analysis is employed at both the individual element and circuit level to establish expected performance metrics for several specific circuit configurations. A protocol based on osmometry is used to experimentally probe mixing behavior in circuits in order to validate these approaches. The overall workflow is applied to two application circuits with immediate use at on the bench-top: series and parallel mixing circuits that are modularly programmable, virtually predictable, highly precise, and operable by hand. PMID:26516059

  9. 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.

  10. 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.

  11. 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.

  12. 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/.

  13. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... NRC's Export Licensing Authority Note: Nuclear fuel elements are manufactured from source or special nuclear material. For oxide fuels, the most common type of fuel equipment for pressing pellets, sintering... the integrity of completed fuel pins (or rods). This item typically includes equipment for: (i)...

  14. Irradiation behavior of ground U(Mo) fuel with and without Si added to the matrix

    NASA Astrophysics Data System (ADS)

    Leenaers, A.; Van den Berghe, S.; Van Renterghem, W.; Charollais, F.; Lemoine, P.; Jarousse, C.; Röhrmoser, A.; Petry, W.

    2011-05-01

    In the framework of the IRIS-TUM irradiation program, several full size, flat dispersion fuel plates containing ground U(Mo) fuel kernels in an aluminum matrix, with and without addition of silicon (2.1 wt.%), have been irradiated in the OSIRIS reactor. The highest irradiated fuel plate (with an Al-Si matrix) reached a local maximum burnup of 88.3% 235U LEU-equivalent and showed a maximum thickness increase of 323 μm (66%) but remained intact. This paper reports the post irradiation examination results obtained on four IRIS-TUM plates. The evolution of the fission gas behavior in this fuel type from homogeneously dispersed nanobubbles to the eventual formation of large but apparently stable fission gas bubbles at the interface of the interaction layer and the fuel kernel is illustrated. It is also shown that the observed moderate, but positive effect of Si as inhibitor for the U(Mo)-Al interaction is related to the dispersion of this element in the interaction layer, although its concentration is very inhomogeneous and appears to be too low to fully inhibit interaction layer growth.

  15. Irradiation behavior of the CNEA's experimental uranium silicide dispersion fuel plates

    SciTech Connect

    Hofman, G.L.; Marajofsky, A.; Kohut, C.; Comision Nacional de Energia Atomica, Buenos Aires )

    1988-01-01

    Since 1978 the CNEA ECBE project has been involved in the development of dispersion fuel plates with four types of fuel materials -- UAl{sub x}, U{sub 3}O{sub 8}, U{sub 3}Si, and U{sub 3}Si{sub 2} -- to be used in low enriched (LEU < 20% {sup 235}U) fuel elements for research reactors. Miniplates with these fuel materials were manufactured at CNEA and were irradiated in the ORR in three series of irradiations as part of the RERTR miniplate irradiation program. The first irradiation contained U{sub 3}O{sub 8} and UAl{sub x} fuel, the second U{sub 3}O{sub 8}, UAl{sub x} and U{sub 3}Si, while the third irradiation test consisted of six U{sub 3}Si{sub 2} miniplates and one U{sub 3}Si miniplate. This third test is the subject of this paper. The present results compare favorably with other irradiations performed in the RERTR program{sup 1,2} showing in particular the excellent behavior of the U{sub 3}Si{sub 2}. The overall data accumulated support the qualification of the CNEA fabrication techniques. 5 refs., 13 figs., 3 tabs.

  16. 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].

  17. Natural convection heat transfer analysis of ATR fuel elements

    SciTech Connect

    Langerman, M.A.

    1992-05-01

    Natural convection air cooling of the Advanced Test Reactor (ATR) fuel assemblies is analyzed to determine the level of decay heat that can be removed without exceeding the melting temperature of the fuel. The study was conducted to assist in the level 2 PRA analysis of a hypothetical ATR water canal draining accident. The heat transfer process is characterized by a very low Rayleigh number (Ra {approx} 10{sup {minus}5}) and a high temperature ratio. Since neither data nor analytical models were available for Ra < 0.1, an analytical approach is presented based upon the integral boundary layer equations. All assumptions and simplifications are presented and assessed and two models are developed from similar foundations. In one model, the well-known Boussinesq approximations are employed, the results from which are used to assess the modeling philosophy through comparison to existing data and published analytical results. In the other model, the Boussinesq approximations are not used, thus making the model more general and applicable to the ATR analysis.

  18. Geochemical behavior of rare earth elements and other trace elements in the Amazon River

    NASA Astrophysics Data System (ADS)

    Merschel, Gila; Bau, Michael; Dantas, Elton Luiz

    2014-05-01

    Rivers transport large amounts of dissolved and suspended particulate material from the catchment area to the oceans and are a major source of trace metals to seawater. The Amazon River is the world's largest river and supplies approximately 20% of the oceans' freshwater (Molinier et al., 1997). However, the behavior of trace elements, especially particle-reactive elements such as the rare earth elements (REE), within the river as well as in the estuary is not well constrained and rather little is known about their transport mechanisms. This study aims at understanding the transport properties of particle-reactive elements in the Amazon River and some of its major tributaries, including the Rio Solimões, Rio Negro, Tapajos, Xingu and Jari Rivers. Samples were taken at 12 stations, seven of which were located in the Amazon mainstream, while the other five stations sampled its tributaries. To account for the effects of variable discharge, the samples were collected during periods of high and low discharge. We present data for major and trace elements, including REE, of the dissolved and suspended load of these samples. First results indicate that the shale-normalized REE pattern of the dissolved load (filtered through 0.2 µm membranes) of the Amazon mainstream and the Rio Solimões confirm earlier studies (Elderfield et al., 1990; Gerard et al., 2003) and show an enrichment of the middle REE relative to the light and heavy REE (LaSN/GdSN: 0.25 - 0.32; GdSN/YbSN: 1.54 - 1.78). In contrast to the Amazon mainstream and the Rio Solimões, which are considered to be whitewater rivers, blackwater rivers, such as the Rio Negro, have a flat REE pattern with higher REE concentrations than whitewater rivers. The third water-type found in the Amazon Basin is clearwater, e.g. Rio Tapajos, with REE patterns in between those of the other two types, i.e. LaSN/GdSN: 0.55 - 0.70; GdSN/YbSN: 1.26 - 1.55. A similar behavior can be identified for other major and trace elements. While

  19. 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.

  20. 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.

  1. LMFBR operational and experimental local-fault experience, primarily with oxide fuel elements

    SciTech Connect

    Warinner, D.K.

    1980-01-01

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS- and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation.

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

    DOEpatents

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

    2013-09-03

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

  3. 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.

  4. 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.

  5. Full-length fuel rod behavior under severe accident conditions

    SciTech Connect

    Lombardo, N J; Lanning, D D; Panisko, F E

    1992-12-01

    This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors.

  6. Application of Phase Field Simulations to Fuel Behavior

    SciTech Connect

    Radhakrishnan, Balasubramaniam; Gorti, Sarma B; Clarno, Kevin T

    2015-01-01

    The application of the phase filed method to simulate the formation and the stress induced re-orientation of zirconium hydride during dry storage of the spent fuel and clad assembly is discussed. The phase field technique is able to capture qualitatively the effect of external stress on the hydride orientation in Zr-H system. However, the modeling effort to-date is far from adequate and several issues remain to be addressed before the simulations can be used as a predictive tool for the behavior of the clad during long time dry storage.

  7. 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.

  8. 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.

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

    SciTech Connect

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

    1997-08-01

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

  10. Transposable elements and small RNAs: Genomic fuel for species diversity

    PubMed Central

    Hoffmann, Federico G; McGuire, Liam P; Counterman, Brian A; Ray, David A

    2015-01-01

    While transposable elements (TE) have long been suspected of involvement in species diversification, identifying specific roles has been difficult. We recently found evidence of TE-derived regulatory RNAs in a species-rich family of bats. The TE-derived small RNAs are temporally associated with the burst of species diversification, suggesting that they may have been involved in the processes that led to the diversification. In this commentary, we expand on the ideas that were briefly touched upon in that manuscript. Specifically, we suggest avenues of research that may help to identify the roles that TEs may play in perturbing regulatory pathways. Such research endeavors may serve to inform evolutionary biologists of the ways that TEs have influenced the genomic and taxonomic diversity around us. PMID:26904375

  11. Nonlinear finite element modeling of dental composite polymerization behavior

    NASA Astrophysics Data System (ADS)

    Laughlin, Gayle A.

    2003-07-01

    Polymerization shrinkage has been one of the primary shortcomings preventing the use of resin composites as a universal dental restorative material. This shrinkage of the bonded restoration causes residual stresses in the composite which in turn are transferred to the adhesive interface. The deleterious effects of this stress environment include compromise of the interface itself and the decrease in the mechanical properties of the cured composite. Novel materials which claim to produce less shrinkage have been presented as a new class of restorative materials that could reduce the effects of this problem. One difficulty in assessing the actual in vivo benefits of these new materials is the fact that there is currently no direct way to measure the stress environment at the composite/tooth clinical interface. Computer modeling using finite element analysis (FEA) could provide helpful information regarding the clinical stress performance of dental composites. The purpose of this study was to develop a model that accurately simulates the nonlinear polymerization behavior of light-cured dental composites using a commercial FEA program, which could be accessible for future research. Two phases were needed to accomplish this purpose. First, a data collection phase included volumetric shrinkage, shrinkage stress, tooth analog strain, and dynamic mechanical analysis experiments. Three composites, a standard methacrylate(Z250) and two experimental low stress epoxy-based composites (oxirane and silorane), were tested. The experimental results revealed an intriguing range of polymerization behavior exhibited by the three composites, indicating that the development of a low stress composite is possible. The information gathered from this phase supplied the necessary material input for the computer modeling, and provided empirical validation data for the model solutions. In the second modeling phase, an FEA approach based on a elastic/viscoplastic material model was used to

  12. 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.

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

    SciTech Connect

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

    2011-10-03

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

  14. Is it possible to model the temperature of the fuel elements in fast reactors using water or air?

    SciTech Connect

    Ushakov, P.A.; Sorokin, A.P.

    1995-12-01

    Thermal stresses caused by temperature nonuniformity around the perimeter of fuel elements in sodium-cooled reactors can cause deformation of the fuel rods. This is the case for fuel elements positioned on the periphery of assemblies and nonuniformly edge-cooled by a coolant, for fuel elements in closely packed assemblies, etc. Extensive investigations of the temperature fields in such fuel elements have been carried out at the Physics and Power Institute of the State Scientific Center, particularly in collaboration with Czech specialists from the Institute of Nuclear Research at Rez. The possibility is now considered of investigating the temperature distribution of fuel elements, for the case when they are closely packed, using test with water and modeling temperature fields when the liquid metals are agitated using tests in air.

  15. An analysis of heating fuel market behavior, 1989--1990

    SciTech Connect

    Not Available

    1990-06-01

    The purpose of this report is to fully assess the heating fuel crisis from a broader and longer-term perspective. Using EIA final, monthly data, in conjunction with credible information from non-government sources, the pricing phenomena exhibited by heating fuels in late December 1989 and early January 1990 are described and evaluated in more detail and more accurately than in the interim report. Additionally, data through February 1990 (and, in some cases, preliminary figures for March) make it possible to assess the market impact of movements in prices and supplies over the heating season as a whole. Finally, the longer time frame and the availability of quarterly reports filed with the Securities and Exchange Commission make it possible to weigh the impact of revenue gains in December and January on overall profits over the two winter quarters. Some of the major, related issues raised during the House and Senate hearings in January concerned the structure of heating fuel markets and the degree to which changes in this structure over the last decade may have influenced the behavior and financial performance of market participants. Have these markets become more concentrated Was collusion or market manipulation behind December's rising prices Did these, or other, factors permit suppliers to realize excessive profits What additional costs were incurred by consumers as a result of such forces These questions, and others, are addressed in the course of this report.

  16. 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.

  17. 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.

  18. 2-D Time-Dependent Fuel Element, Thermal Analysis Code System.

    2001-09-24

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

  19. Fuel models to predict fire behavior in untreated conifer slash. Forest Service research note (final)

    SciTech Connect

    Salazar, L.A.; Bevins, C.D.

    1984-11-01

    Fire behavior in untreated slash of nine conifer species was simulated for 10 successive years after logging. Two aging factors that affect fire behavior--fuel bed compaction and foliage retention--were modeled by least squares regression techniques. On the basis of spread rate and flame length for a set of weather observations, standard Northern Forest Fire Laboratory fuel models were chosen to predict fire behavior for aging slash of each species at three initial fuel loadings. Differences in the standard fuel model sequences best representing aging process among species were most influenced by foliage surface-area-to-volume ratio, and such differences increased as initial fuel load increased.

  20. High temperature fuel/emitter system for advanced thermionic fuel elements

    NASA Astrophysics Data System (ADS)

    Moeller, Helen H.; Bremser, Albert H.; Gontar, Alexander; Fiviesky, Evgeny

    1997-01-01

    Specialists in space applications are currently focusing on bimodal power systems designed to provide both electric power and thermal propulsion (Kennedy, 1994 and Houts, 1995). Our work showed that thermionics is a viable technology for nuclear bimodal power systems. We demonstrated that materials for a thermionic fuel-emitter combination capable of performing at operating temperatures of 2473 K are not only possible but available. The objective of this work, funded by the US Department of Energy, Office of Space and Defense Power Systems, was to evaluate the compatibility of fuel material consisting of an uranium carbide/tantalum carbide solid solution with an emitter material consisting of a monocrystalline tungsten-niobium alloy. The uranium loading of the fuel material was 70 mole% uranium carbide. The program was successfully accomplished by a B&W/SIA LUTCH team. Its workscope was integrated with tasks being performed at both Babcock & Wilcox, Lynchburg Research Center, Lynchburg, Virginia, and SIA LUTCH, Podolsk, Russia. Samples were fabricated by LUTCH and seven thermal tests were performed in a hydrogen atmosphere. The first preliminary test was performed at 2273 K by LUTCH, and the remaining six tests were performed At B&W. Three tests were performed at 2273 K, two at 2373 K, and the final test at 2473 K. The results showed that the fuel and emitter materials were compatible in the presence of hydrogen. No evidence of liquid formation, dissolution of the uranium carbide from the uranium carbide/tantalum carbide solid solution, or diffusion of the uranium into the monocrystalline tungsten alloy was observed. Among the highlights of the program was the successful export of the fuel samples from Russia and their import into the US by commercial transport. This paper will discuss the technical aspects of this work.

  1. Development of customized fire behavior fuel models for boreal forests of northeastern China.

    PubMed

    Wu, Zhi Wei; He, Hong Shi; Chang, Yu; Liu, Zhi Hua; Chen, Hong Wei

    2011-12-01

    Knowledge of forest fuels and their potential fire behavior across a landscape is essential in fire management. Four customized fire behavior fuel models that differed significantly in fuels characteristics and environmental conditions were identified using hierarchical cluster analysis based on fuels data collected across a boreal forest landscape in northeastern China. Fuel model I represented the dense and heavily branched Pinus pumila shrubland which has significant fine live woody fuels. These forests occur mainly at higher mountain elevations. Fuel model II is applicable to forests dominated by Betula platyphylla and Populus davidiana occurring in native forests on hill slopes or at low mountain elevations. This fuel model was differentiated from other fuel models by higher herbaceous cover and lower fine live woody loading. The primary coniferous forests dominated by Larix gmelini and Pinus sylvestris L. var. mongolica were classified as fuel model III and fuel model IV. Those fuel models differed from one another in average cover and height of understory shrub and herbaceous layers as well as in aspect. The potential fire behavior for each fuel model was simulated with the BehavePlus5.0 fire behavior prediction system. The simulation results indicated that the Pinus pumila shrubland fuels had the most severe fire behavior for the 97th percentile weather condition, and had the least severe fire behavior under 90th percentile weather condition. Fuel model II presented the least severe fire potential across weather conditions. Fuel model IV resulted in greater fire severity than Fuel model III across the two weather scenarios that were examined.

  2. 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.

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

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-03

    ... Information The NRC published DG-2005 in the Federal Register on March 22, 2012 (77 FR 16868) for a 60-day... COMMISSION Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test... Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This...

  5. 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

  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. Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

    NASA Astrophysics Data System (ADS)

    Kwon, Young Joo; Choi, Jong Won

    This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

  8. Development of custom fire behavior fuel models from FCCS fuelbeds for the Savannah River fuel assessment project.

    SciTech Connect

    Scott, Joe, H.

    2009-07-23

    The purpose of this project is to create fire behavior fuel models that replicate the fire behavior characteristics (spread rate and fireline intensity) produced by 23 candidate FCCS fuelbeds developed for the Savannah River National Wildlife Refuge. These 23 fuelbeds were created by FERA staff in consultation with local fuel managers. The FCCS produces simulations of surface fire spread rate and flame length (and therefore fireline intensity) for each of these fuelbeds, but it does not produce maps of those fire behavior characteristics or simulate fire growth—those tasks currently require the use of the FARSITE and/or FlamMap software systems. FARSITE and FlamMap do not directly use FCCS fuelbeds, but instead use standard or custom fire behavior fuel models to describe surface fuel characteristics for fire modeling. Therefore, replicating fire growth and fire behavior potential calculations using FCCS-simulated fire characteristics requires the development of custom fuel models that mimic, as closely as possible, the fire behavior characteristics produced by the FCCS for each fuelbed, over a range of fuel moisture and wind speeds.

  9. Dynamic behavior of PEM fuel cell and microturbine power plants

    NASA Astrophysics Data System (ADS)

    El-Sharkh, M. Y.; Sisworahardjo, N. S.; Uzunoglu, M.; Onar, O.; Alam, M. S.

    This paper presents a comparison between the dynamic behavior of a 250 kW stand-alone proton exchange membrane fuel cell power plant (PEM FCPP) and a 250 kW stand-alone microturbine (MT). Dynamic models for the two are introduced. To control the voltage and the power output of the PEM FCPP, voltage and power control loops are added to the model. For the MT, voltage, speed, and power control are used. Dynamic models are used to determine the response of the PEM FCPP and MT to a load step change. Simulation results indicate that the response of the MT to reach a steady state is about twice as fast as the PEM FCPP. For stand-alone operation of a PEM FCPP, a set of batteries or ultracapacitors is needed in order to satisfy the power mismatch during transient periods. Software simulation results are obtained by using MATLAB ®, Simulink ®, and SimPowerSystems ®.

  10. 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.

  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. 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.

  13. Fabrication of simulated plate fuel elements: Defining role of stress relief annealing

    NASA Astrophysics Data System (ADS)

    Kohli, D.; Rakesh, R.; Sinha, V. P.; Prasad, G. J.; Samajdar, I.

    2014-04-01

    This study involved fabrication of simulated plate fuel elements. Uranium silicide of actual fuel elements was replaced with yttria. The fabrication stages were otherwise identical. The final cold rolled and/or straightened plates, without stress relief, showed an inverse relationship between bond strength and out of plane residual shear stress (τ13). Stress relief of τ13 was conducted over a range of temperatures/times (200-500 °C and 15-240 min) and led to corresponding improvements in bond strength. Fastest τ13 relief was obtained through 300 °C annealing. Elimination of microscopic shear bands, through recovery and partial recrystallization, was clearly the most effective mechanism of relieving τ13.

  14. Fabrication of simulated plate fuel elements: Defining role of out-of-plane residual shear stress

    NASA Astrophysics Data System (ADS)

    Rakesh, R.; Kohli, D.; Sinha, V. P.; Prasad, G. J.; Samajdar, I.

    2014-02-01

    Bond strength and microstructural developments were investigated during fabrication of simulated plate fuel elements. The study involved roll bonding of aluminum-aluminum (case A) and aluminum-aluminum + yttria (Y2O3) dispersion (case B). Case B approximated aluminum-uranium silicide (U3Si2) 'fuel-meat' in an actual plate fuel. Samples after different stages of fabrication, hot and cold rolling, were investigated through peel and pull tests, micro-hardness, residual stresses, electron and micro-focus X-ray diffraction. Measurements revealed a clear drop in bond strength during cold rolling: an observation unique to case B. This was related to significant increase in 'out-of-plane' residual shear stresses near the clad/dispersion interface, and not from visible signatures of microstructural heterogeneities.

  15. Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

    2002-08-01

    Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

  16. Concrete shielding housing for receiving and storing a nuclear fuel element container

    SciTech Connect

    Dyck, H.-P.

    1985-07-02

    The invention is directed to a concrete shielding housing for receiving and storing a fuel element container filled with spent nuclear reactor fuel elements. The container is suitable for transport and storage. The clear interior dimensions of the concrete shielding housing are somewhat larger than the outer dimensions of the fuel element container. The concrete shielding housing includes a pallet-type base and in the lower region of the housing there is provided at least one air inlet opening and in the upper region of the housing there is provided at least one air outlet opening. To prevent an uncontrolled conduction of moisture away from the interior of the housing to the ground or to the floor of a storage area or building, there is provided a collection pan arranged under the base plate of the pallet-like base. At least one axial bore extends clear through the base plate of the pallet-like base. With the arrangement of the collection pan, contaminated moisture is collected and prevented from seeping into the ground or floor.

  17. 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.

  18. Developing Custom Fire Behavior Fuel Models for Mediterranean Wildland-Urban Interfaces in Southern Italy.

    PubMed

    Elia, Mario; Lafortezza, Raffaele; Lovreglio, Raffaella; Sanesi, Giovanni

    2015-09-01

    The dramatic increase of fire hazard in wildland-urban interfaces (WUIs) has required more detailed fuel management programs to preserve ecosystem functions and human settlements. Designing effective fuel treatment strategies allows to achieve goals such as resilient landscapes, fire-adapted communities, and ecosystem response. Therefore, obtaining background information on forest fuel parameters and fuel accumulation patterns has become an important first step in planning fuel management interventions. Site-specific fuel inventory data enhance the accuracy of fuel management planning and help forest managers in fuel management decision-making. We have customized four fuel models for WUIs in southern Italy, starting from forest classes of land-cover use and adopting a hierarchical clustering approach. Furthermore, we provide a prediction of the potential fire behavior of our customized fuel models using FlamMap 5 under different weather conditions. The results suggest that fuel model IIIP (Mediterranean maquis) has the most severe fire potential for the 95th percentile weather conditions and the least severe potential fire behavior for the 85th percentile weather conditions. This study shows that it is possible to create customized fuel models directly from fuel inventory data. This achievement has broad implications for land managers, particularly forest managers of the Mediterranean landscape, an ecosystem that is susceptible not only to wildfires but also to the increasing human population and man-made infrastructures. PMID:25962800

  19. Developing Custom Fire Behavior Fuel Models for Mediterranean Wildland-Urban Interfaces in Southern Italy

    NASA Astrophysics Data System (ADS)

    Elia, Mario; Lafortezza, Raffaele; Lovreglio, Raffaella; Sanesi, Giovanni

    2015-09-01

    The dramatic increase of fire hazard in wildland-urban interfaces (WUIs) has required more detailed fuel management programs to preserve ecosystem functions and human settlements. Designing effective fuel treatment strategies allows to achieve goals such as resilient landscapes, fire-adapted communities, and ecosystem response. Therefore, obtaining background information on forest fuel parameters and fuel accumulation patterns has become an important first step in planning fuel management interventions. Site-specific fuel inventory data enhance the accuracy of fuel management planning and help forest managers in fuel management decision-making. We have customized four fuel models for WUIs in southern Italy, starting from forest classes of land-cover use and adopting a hierarchical clustering approach. Furthermore, we provide a prediction of the potential fire behavior of our customized fuel models using FlamMap 5 under different weather conditions. The results suggest that fuel model IIIP (Mediterranean maquis) has the most severe fire potential for the 95th percentile weather conditions and the least severe potential fire behavior for the 85th percentile weather conditions. This study shows that it is possible to create customized fuel models directly from fuel inventory data. This achievement has broad implications for land managers, particularly forest managers of the Mediterranean landscape, an ecosystem that is susceptible not only to wildfires but also to the increasing human population and man-made infrastructures.

  20. Developing Custom Fire Behavior Fuel Models for Mediterranean Wildland-Urban Interfaces in Southern Italy.

    PubMed

    Elia, Mario; Lafortezza, Raffaele; Lovreglio, Raffaella; Sanesi, Giovanni

    2015-09-01

    The dramatic increase of fire hazard in wildland-urban interfaces (WUIs) has required more detailed fuel management programs to preserve ecosystem functions and human settlements. Designing effective fuel treatment strategies allows to achieve goals such as resilient landscapes, fire-adapted communities, and ecosystem response. Therefore, obtaining background information on forest fuel parameters and fuel accumulation patterns has become an important first step in planning fuel management interventions. Site-specific fuel inventory data enhance the accuracy of fuel management planning and help forest managers in fuel management decision-making. We have customized four fuel models for WUIs in southern Italy, starting from forest classes of land-cover use and adopting a hierarchical clustering approach. Furthermore, we provide a prediction of the potential fire behavior of our customized fuel models using FlamMap 5 under different weather conditions. The results suggest that fuel model IIIP (Mediterranean maquis) has the most severe fire potential for the 95th percentile weather conditions and the least severe potential fire behavior for the 85th percentile weather conditions. This study shows that it is possible to create customized fuel models directly from fuel inventory data. This achievement has broad implications for land managers, particularly forest managers of the Mediterranean landscape, an ecosystem that is susceptible not only to wildfires but also to the increasing human population and man-made infrastructures.

  1. 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.

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

    SciTech Connect

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

    1993-01-20

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

  3. Dynamic behavior of gasoline fuel cell electric vehicles

    NASA Astrophysics Data System (ADS)

    Mitchell, William; Bowers, Brian J.; Garnier, Christophe; Boudjemaa, Fabien

    As we begin the 21st century, society is continuing efforts towards finding clean power sources and alternative forms of energy. In the automotive sector, reduction of pollutants and greenhouse gas emissions from the power plant is one of the main objectives of car manufacturers and innovative technologies are under active consideration to achieve this goal. One technology that has been proposed and vigorously pursued in the past decade is the proton exchange membrane (PEM) fuel cell, an electrochemical device that reacts hydrogen with oxygen to produce water, electricity and heat. Since today there is no existing extensive hydrogen infrastructure and no commercially viable hydrogen storage technology for vehicles, there is a continuing debate as to how the hydrogen for these advanced vehicles will be supplied. In order to circumvent the above issues, power systems based on PEM fuel cells can employ an on-board fuel processor that has the ability to convert conventional fuels such as gasoline into hydrogen for the fuel cell. This option could thereby remove the fuel infrastructure and storage issues. However, for these fuel processor/fuel cell vehicles to be commercially successful, issues such as start time and transient response must be addressed. This paper discusses the role of transient response of the fuel processor power plant and how it relates to the battery sizing for a gasoline fuel cell vehicle. In addition, results of fuel processor testing from a current Renault/Nuvera Fuel Cells project are presented to show the progress in transient performance.

  4. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    SciTech Connect

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  5. NASA Human Research Program: Behavioral Health and Performance Program Element

    NASA Technical Reports Server (NTRS)

    Leveton, Lauren B.

    2009-01-01

    This viewgraph presentation reviews the performance errors associated with sleep loss, fatigue and psychomotor factors during manned space flight. Short and long term behavioral health factors are also addressed

  6. Experience with failed LMR oxide fuel element performance in European fast reactors

    NASA Astrophysics Data System (ADS)

    Plitz, H.; Crittenden, G. C.; Languille, A.

    1993-09-01

    The performance of failed fuel has great significance for the safe and economic operation of LMR's, and considerable experience has accrued from experimental defect pin irradiations and naturally occurring failures in European test and prototype reactors. To data 60 natural fuel element failures have been recorded in PFR, Phénix and KNK II, 41 with exposed fuel and 19 as gas leakers. The various failures occurred during all stages of pin lifetimes, i.e. at the very beginning (0.3 at% burn-up) as well as at medium and at very high burn-up. The present experience extends up to 190 GWd/t and up to 135 dpaNRT. Based on the experience we can state: (i) Even large defects at end-of-life pins resulted in limited fuel loss (ii) No pin-to-pin failure propagation has been observed (iii) The reaction produces formed by the chemical reaction sodium/mixed oxide and the kinetics act beneficially and may protect open cracks. For the European Fast Reactor (EFR) project additional work is being performed, with regard to the EFR requirements of pin design (covering normal operation and incidental events) and the behaviour of failed pins under storage conditions.

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

    SciTech Connect

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

    2001-11-06

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

  8. Performance optimization considerations for thermionic fuel elements in a heat pipe cooled thermionic reactor

    NASA Astrophysics Data System (ADS)

    Bellis, Elizabeth A.

    1992-01-01

    A heat pipe-cooled, in-core thermionic (HPTI) reactor design has been proposed in support of the Air Force Thermionic Space Nuclear Power Program. As part of this design, the performance of the power conversion system has been characterized. This paper focuses on the performance optimization studies carried out of a thermionic fuel element (TFE) which will be used in a reactor design capable of producing 40 kWe over a 10 year operating life. The technical approach to the optimization studies closely couples converter lifetime constraints with converter performance to produce the best possible design choice.

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

    SciTech Connect

    SCHWINKENDORF, K.N.

    2006-05-12

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

  10. Fuel Regression Rate Behavior of CAMUI Hybrid Rocket

    NASA Astrophysics Data System (ADS)

    Kaneko, Yudai; Itoh, Mitsunori; Kakikura, Akihito; Mori, Kazuhiro; Uejima, Kenta; Nakashima, Takuji; Wakita, Masashi; Totani, Tsuyoshi; Oshima, Nobuyuki; Nagata, Harunori

    A series of static firing tests was conducted to investigate the fuel regression characteristics of a Cascaded Multistage Impinging-jet (CAMUI) type hybrid rocket motor. A CAMUI type hybrid rocket uses the combination of liquid oxygen and a fuel grain made of polyethylene as a propellant. The collision distance divided by the port diameter, H/D, was varied to investigate the effect of the grain geometry on the fuel regression rate. As a result, the H/D geometry has little effect on the regression rate near the stagnation point, where the heat transfer coefficient is high. On the contrary, the fuel regression rate decreases near the circumference of the forward-end face and the backward-end face of fuel blocks. Besides the experimental approaches, a method of computational fluid dynamics clarified the heat transfer distribution on the grain surface with various H/D geometries. The calculation shows the decrease of the flow velocity due to the increase of H/D on the area where the fuel regression rate decreases with the increase of H/D. To estimate the exact fuel consumption, which is necessary to design a fuel grain, real-time measurement by an ultrasonic pulse-echo method was performed.

  11. Developing custom fire behavior fuel models from ecologically complex fuel structures for upper Atlantic Coastal Plain forests.

    SciTech Connect

    Parresol, Bernard, R.; Scott, Joe, H.; Andreu, Anne; Prichard, Susan; Kurth, Laurie

    2012-01-01

    Currently geospatial fire behavior analyses are performed with an array of fire behavior modeling systems such as FARSITE, FlamMap, and the Large Fire Simulation System. These systems currently require standard or customized surface fire behavior fuel models as inputs that are often assigned through remote sensing information. The ability to handle hundreds or thousands of measured surface fuelbeds representing the fine scale variation in fire behavior on the landscape is constrained in terms of creating compatible custom fire behavior fuel models. In this study, we demonstrate an objective method for taking ecologically complex fuelbeds from inventory observations and converting those into a set of custom fuel models that can be mapped to the original landscape. We use an original set of 629 fuel inventory plots measured on an 80,000 ha contiguous landscape in the upper Atlantic Coastal Plain of the southeastern United States. From models linking stand conditions to component fuel loads, we impute fuelbeds for over 6000 stands. These imputed fuelbeds were then converted to fire behavior parameters under extreme fuel moisture and wind conditions (97th percentile) using the fuel characteristic classification system (FCCS) to estimate surface fire rate of spread, surface fire flame length, shrub layer reaction intensity (heat load), non-woody layer reaction intensity, woody layer reaction intensity, and litter-lichen-moss layer reaction intensity. We performed hierarchical cluster analysis of the stands based on the values of the fire behavior parameters. The resulting 7 clusters were the basis for the development of 7 custom fire behavior fuel models from the cluster centroids that were calibrated against the FCCS point data for wind and fuel moisture. The latter process resulted in calibration against flame length as it was difficult to obtain a simultaneous calibration against both rate of spread and flame length. The clusters based on FCCS fire behavior

  12. Fuel type characterization and potential fire behavior estimation in Sardinia and Corsica islands

    NASA Astrophysics Data System (ADS)

    Bacciu, V.; Pellizzaro, G.; Santoni, P.; Arca, B.; Ventura, A.; Salis, M.; Barboni, T.; Leroy, V.; Cancellieri, D.; Leoni, E.; Ferrat, L.; Perez, Y.; Duce, P.; Spano, D.

    2012-04-01

    Wildland fires represent a serious threat to forests and wooded areas of the Mediterranean Basin. As recorded by the European Commission (2009), during the last decade Southern Countries have experienced an annual average of about 50,000 forest fires and about 470,000 burned hectares. The factor that can be directly manipulated in order to minimize fire intensity and reduce other fire impacts, such as three mortality, smoke emission, and soil erosion, is wildland fuel. Fuel characteristics, such as vegetation cover, type, humidity status, and biomass and necromass loading are critical variables in affecting wildland fire occurrence, contributing to the spread, intensity, and severity of fires. Therefore, the availability of accurate fuel data at different spatial and temporal scales is needed for fire management applications, including fire behavior and danger prediction, fire fighting, fire effects simulation, and ecosystem simulation modeling. In this context, the main aims of our work are to describe the vegetation parameters involved in combustion processes and develop fire behavior fuel maps. The overall work plan is based firstly on the identification and description of the different fuel types mainly affected by fire occurrence in Sardinia (Italy) and Corsica (France) Islands, and secondly on the clusterization of the selected fuel types in relation to their potential fire behavior. In the first part of the work, the available time series of fire event perimeters and the land use map data were analyzed with the purpose of identifying the main land use types affected by fires. Thus, field sampling sites were randomly identified on the selected vegetation types and several fuel variables were collected (live and dead fuel load partitioned following Deeming et al., (1977), depth of fuel layer, plant cover, surface area-to-volume ratio, heat content). In the second part of the work, the potential fire behavior for every experimental site was simulated using

  13. NASA Human Research Program Behavioral Health and Performance Element (BHP)

    NASA Technical Reports Server (NTRS)

    Whitmire, Sandra; Faulk, Jeremy; Leveton, Lauren

    2010-01-01

    The goal of NASA BHP is to identify, characterize, and prevent or reduce behavioral health and performance risks associated with space travel, exploration, and return to terrestrial life. The NASA Behavioral Health and Performance Operations Group (BHP Ops) supports astronauts and their families before, during, and after a long-duration mission (LDM) on the ISS. BHP Ops provides ISS crews with services such as preflight training (e.g., psychological factors of LDM, psychological support, cross-cultural); preflight, in-flight, and postflight support services, including counseling for astronauts and their families; and psychological support such as regular care packages and a voice-over IP phone system between crew members and their families to facilitate real-time one-on-one communication.

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

    SciTech Connect

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

    1981-11-01

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

  15. Finite element analysis of notch tensile behavior of alloy 718

    NASA Astrophysics Data System (ADS)

    Sridhar, A.; Srivathsa, B.

    2013-06-01

    Notch tensile behavior of alloy 718 is characterized in conventionally heat treated condition as a function of U and V notches at 25, 200 & 400 °C. The experimental results were then compared with the values obtained from simulation of notched geometries in ANSYS software using smooth specimen data. An excellent agreement is noticed between simulated and experimental true stress-true strain curves.

  16. Elemental characterization of particulate matter emitted from biomass burning: Wind tunnel derived source profiles for herbaceous and wood fuels

    NASA Astrophysics Data System (ADS)

    Turn, S. Q.; Jenkins, B. M.; Chow, J. C.; Pritchett, L. C.; Campbell, D.; Cahill, T.; Whalen, S. A.

    1997-02-01

    Particulate matter emitted from wind tunnel simulations of biomass burning for five herbaceous crop residues (rice, wheat and barley straws, corn stover, and sugar cane trash) and four wood fuels (walnut and almond prunings and ponderosa pine and Douglas fir slash) was collected and analyzed for major elements and water soluble species. Primary constituents of the particulate matter were C, K, Cl, and S. Carbon accounted for roughly 50% of the herbaceous fuel PM and about 70% for the wood fuels. For the herbaceous fuels, particulate matter from rice straw in the size range below 10 μm aerodynamic diameter (PM10) had the highest concentrations of both K (24%) and Cl, (17%) and barley straw PM10 contained the highest sulfur content (4%). K, Cl, and S were present in the PM of the wood fuels at reduced levels with maximum concentrations of 6.5% (almond prunings), 3% (walnut prunings), and 2% (almond prunings), respectively. Analysis of water soluble species indicated that ionic forms of K, Cl, and S made up the majority of these elements from all fuels. Element balances showed K, Cl, S, and N to have the highest recovery factors (fraction of fuel element found in the particulate matter) in the PM of the elements analyzed. In general, chlorine was the most efficiently recovered element for the herbaceous fuels (10 to 35%), whereas sulfur recovery was greatest for the wood fuels (25 to 45%). Unique potassium to elemental carbon ratios of 0.20 and 0.95 were computed for particulate matter (PM10 K/C(e)) from herbaceous and wood fuels, respectively. Similarly, in the size class below 2.5 μm, high-temperature elemental carbon to bromine (PM2.5 C(eht)/Br) ratios of ˜7.5, 43, and 150 were found for the herbaceous fuels, orchard prunings, and forest slash, respectively. The molar ratios of particulate phase bromine to gas phase CO2 (PM10 Br/CO2) are of the same order of magnitude as gas phase CH3Br/CO2 reported by others.

  17. Beyond FIRO-B--three new theory-derived measures--Element B: behavior, Element F: feelings, Element S: self.

    PubMed

    Schutz, W

    1992-06-01

    Although the FIRO-B instrument has been used widely for a large number of purposes, it was not designed as a general purpose instrument. Several years ago, after revising the FIRO theory underlying the instrument based on over 20 years' experience with the instrument and related activities, the author revised the FIRO-B extensively, so extensively it was given a new name, Element B. The new instrument is much stronger both theoretically and psychometrically while at the same time retaining the simplicity and shortness of the original. In addition, two new instruments based on the same theory were designed, developed, and tested. They measure feelings (Element F) and self-concept (Element S). All three instruments have, over the past 10 years, been used primarily as training instruments. When given in conjunction with other methods, they have been used for improving self-awareness, teamwork, morale, and productivity in such organizations as Procter & Gamble, AT&T, NASA, Amdahl Corporation, the Swedish Army, and about 100 companies in Japan. Included is a comment on scales anchored both logically, using methods such as facet design and unidimensional scaling, and empirically, such as the "big five."

  18. Physical and chemical behavior of flowing endothermic jet fuels

    NASA Astrophysics Data System (ADS)

    Ward, Thomas Arthur

    Hydrocarbon fuels have been used as cooling media for aircraft jet engines for decades. However, modern aircraft engines are reaching a practical heat transfer limit beyond which the convective heat transfer provided by fuels is no longer adequate. One solution is to use an endothermic fuel that absorbs heat through a series of pyrolytic chemical reactions. However, many of the physical and chemical processes involved in endothermic fuel degradation are not well understood. The purpose of this dissertation is to study different characteristics of endothermic fuels using experiments and computational models. In the first section, data from three flow experiments using heated Jet-A fuel and additives were analyzed (with the aid of CFD calculations) to study the effects of treated surfaces on surface deposition. Surface deposition is the primary impediment in creating an operational endothermic fuel heat exchanger system, because deposits can obstruct fuel pathways causing a catastrophic system failure. As heated fuel flows through a fuel system, trace species within the fuel react with dissolved O2 to form surface deposits. At relatively higher fuel temperatures, the dissolved O2 is depleted, and pyrolytic chemistry becomes dominant (at temperatures greater than ˜500 °C). In the first experiment, the dissolved O2 consumption of heated fuel was measured on different surface types over a range of temperatures. It is found that use of treated tubes significantly delays oxidation of the fuel. In the second experiment, the treated length of tubing was progressively increased, which varied the characteristics of the thermal-oxidative deposits formed. In the third experiment, pyrolytic surface deposition in either fully treated or untreated tubes is studied. It is found that the treated surface significantly reduced the formation of surface deposits for both thermal oxidative and pyrolytic degradation mechanisms. Moreover, it is found that the chemical reactions resulting

  19. 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

  20. 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.

  1. Lanthanides in Metallic Nuclear Fuels: Their Behavior and Methods for Their Control

    SciTech Connect

    Robert D. Mariani; Douglas L. Porter; Thomas P. O'Holleran; Steven L. Hayes; J. Rory Kennedy

    2011-12-01

    The thermodynamic and experimental basis is given for using dopant additives to bind lanthanides as intermetallic compounds in metallic nuclear fuels. Lanthanide fission products are a major factor in limiting the lifetime of the fuel, because they migrate to the fuel slug peripheral surface where they participate in fuel-cladding chemical interactions (FCCI) with the steel cladding. Lanthanide carryover in recycled metal fuels can accelerate FCCI, as recycled lanthanides would likely segregate from the fuel phase, putting the lanthanides in prompt contact with the cladding. In out-of-pile tests we examined the use of Pd for binding the lanthanides, with Pd selected because of its known metallurgical properties in fuel related systems and because of its known behavior in irradiated EBR-II fuels. Initial results confirmed that palladium may be expected to mitigate FCCI arising from lanthanides, and it has been recommended for in-pile tests. We also evaluated transport phenomena responsible for lanthanide migration, and identified liquid-like behaviors as being dominant. Liquid-like behaviors include transport with liquid metals, liquid metal solutions, and rapid surface transport of alloys/metals near their melting temperatures. The analysis led to establishing general criteria for selecting alternate dopant additives, and identifying Sn, Sb, and Te as alternates for further testing.

  2. Enhancing the ABAQUS Thermomechanics Code to Simulate Steady and Transient Fuel Rod Behavior

    SciTech Connect

    R. L. Williamson; D. A. Knoll

    2009-09-01

    A powerful multidimensional fuels performance capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth , gap heat transfer, and gap/plenum gas behavior during irradiation. The various modeling capabilities are demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multi-pellet fuel rod, during both steady and transient operation. Computational results demonstrate the importance of a multidimensional fully-coupled thermomechanics treatment. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermo-mechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

  3. Discrete Element study of granular material - Bumpy wall interface behavior

    NASA Astrophysics Data System (ADS)

    El Cheikh, Khadija; Rémond, Sébastien; Pizette, Patrick; Vanhove, Yannick; Djelal, Chafika

    2016-09-01

    This paper presents a DEM study of a confined granular material sheared between two parallel bumpy walls. The granular material consists of packed dry spherical particles. The bumpiness is modeled by spheres of a given diameter glued on horizontal planes. Different bumpy surfaces are modeled by varying diameter or concentration of glued spheres. The material is sheared by moving the two bumpy walls at fixed velocity. During shear, the confining pressure applied on each bumpy wall is controlled. The effect of wall bumpiness on the effective friction coefficient and on the granular material behavior at the bumpy walls is reported for various shearing conditions. For given bumpiness and confining pressure that we have studied, it is found that the shear velocity does not affect the shear stress. However, the effective friction coefficient and the behavior of the granular material depend on the bumpiness. When the diameter of the glued spheres is larger than about the average grains diameter of the medium, the latter is uniformly sheared and the effective friction coefficient remains constant. For smaller diameters of the glued spheres, the effective friction coefficient increases with the diameter of glued spheres. The influence of glued spheres concentration is significant only for small glued spheres diameters, typically half of average particle diameter of the granular material. In this case, increasing the concentration of glued spheres leads to a decrease in effective friction coefficient and to shear localization at the interface. For different diameters and concentrations of glued spheres, we show that the effect of bumpiness on the effective friction coefficient can be characterized by the depth of interlocking.

  4. Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

    SciTech Connect

    Weber, J.W.

    1994-09-28

    The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

  5. A physical description of fission product behavior fuels for advanced power reactors.

    SciTech Connect

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

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

    PubMed

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

    2016-01-01

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

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

    PubMed

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

    2016-01-01

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

  8. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    NASA Astrophysics Data System (ADS)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  9. Understanding selected trace elements behavior in a coal-fired power plant in Malaysia for assessment of abatement technologies.

    PubMed

    Mokhtar, Mutahharah M; Taib, Rozainee M; Hassim, Mimi H

    2014-08-01

    The Proposed New Environmental Quality (Clean Air) Regulation 201X (Draft), which replaces the Malaysia Environmental Quality (Clean Air) 1978, specifies limits to additional pollutants from power generation using fossil fuel. The new pollutants include Hg, HCl, and HF with limits of 0.03, 100, and 15 mg/N-m3 at 6% O2, respectively. These pollutants are normally present in very small concentrations (known as trace elements [TEs]), and hence are often neglected in environmental air quality monitoring in Malaysia. Following the enactment of the new regulation, it is now imperative to understand the TEs behavior and to assess the capability of the existing abatement technologies to comply with the new emission limits. This paper presents the comparison of TEs behavior of the most volatile (Hg, Cl, F) and less volatile (As, Be, Cd, Cr, Ni, Se, Pb) elements in subbituminous and bituminous coal and coal combustion products (CCP) (i.e., fly ash and bottom ash) from separate firing of subbituminous and bituminous coal in a coal-fired power plant in Malaysia. The effect of air pollution control devices configuration in removal of TEs was also investigated to evaluate the effectiveness of abatement technologies used in the plant. This study showed that subbituminous and bituminous coals and their CCPs have different TEs behavior. It is speculated that ash content could be a factor for such diverse behavior In addition, the type of coal and the concentrations of TEs in feed coal were to some extent influenced by the emission of TEs in flue gas. The electrostatic precipitator (ESP) and seawater flue gas desulfurization (FGD) used in the studied coal-fired power plant were found effective in removing TEs in particulate and vapor form, respectively, as well as complying with the new specified emission limits. Implications: Coals used by power plants in Peninsular Malaysia come from the same supplier (Tenaga Nasional Berhad Fuel Services), which is a subsidiary of the Malaysia

  10. Effective elements of school health promotion across behavioral domains: a systematic review of reviews

    PubMed Central

    Peters, Louk WH; Kok, Gerjo; Ten Dam, Geert TM; Buijs, Goof J; Paulussen, Theo GWM

    2009-01-01

    Background Most school health education programs focus on a single behavioral domain. Integrative programs that address multiple behaviors may be more efficient, but only if the elements of change are similar for these behaviors. The objective of this study was to examine which effective elements of school health education are similar across three particular behavioral domains. Methods A systematic review of reviews of the effectiveness of school-based health promotion programs was conducted for the domains of substance abuse, sexual behavior, and nutrition. The literature search spanned the time period between 1995 and October 2006 and included three databases, websites of review centers and backward search. Fifty-five reviews and meta-analyses met predetermined relevance and publication criteria and were included. Data was extracted by one reviewer and checked by a second reviewer. A standardized data extraction form was used, with detailed attention to effective elements pertaining to program goals, development, content, methods, facilitator, components and intensity. Two assessors rated the quality of reviews as strong, moderate or weak. We included only strong and moderate reviews in two types of analysis: one based on interpretation of conflicting results, the other on a specific vote-counting rule. Results Thirty six reviews were rated strong, 6 moderate, and 13 weak. A multitude of effective elements was identified in the included reviews and many elements were similar for two or more domains. In both types of analysis, five elements with evidence from strong reviews were found to be similar for all three domains: use of theory; addressing social influences, especially social norms; addressing cognitive-behavioral skills; training of facilitators; and multiple components. Two additional elements had positive results in all domains with the rule-based method of analysis, but had inconclusive results in at least one domain with the interpretion-based method

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

    SciTech Connect

    TOFFER, H.

    2006-07-18

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

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

    SciTech Connect

    Tsourides, D.

    1998-12-31

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

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

    SciTech Connect

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

    1996-03-01

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

  14. Performance simulation of an advanced cylindrical thermionic fuel element with a graphite reservoir

    NASA Astrophysics Data System (ADS)

    Young, Timothy J.; Thayer, Kevin L.; Ramalingam, Mysore L.

    1993-07-01

    This paper describes the analytical work to optimize the steady-state electrical and thermal characteristics of an advanced, power producing, cylindrical thermionic fuel element (TFE) operating in a space nuclear reactor. The thermionic converter was equipped with an integral, lamellar graphite-cesium reservoir attached to the non-nuclear fueled TFE emitter lead as a means for supplying cesium vapor for efficient thermionic emission. Five intercalated cesium-graphite compounds were chosen for this analysis and the optimum position for the placement of each candidate reservoir in the TFE lead region was determined. The Advanced Thermionic Initiative (ATI) thermal spectrum, 'driverless' nuclear reactor configuration, with an output of 36 kWe, was used as a basis for the calculations. A coupled thermionic and thermal-hydraulic computer program was integrated with a lead region thermal model to calculate the thermal and electrical output characteristics of the TFE for different reservoir locations. The results of this analysis indicate that the temperature distribution in the lead region of the TFE at steady-state is such that only four of the candidate reservoirs analyzed could be located on the lead and supply the requisite cesium vapor pressure for optimum TFE operation.

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

    DOEpatents

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

    2011-12-13

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

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

    NASA Astrophysics Data System (ADS)

    Forquin, P.

    2010-06-01

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

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

    SciTech Connect

    Wernsman, B.

    1997-01-01

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

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

    SciTech Connect

    Wernsman, Bernard

    1997-01-10

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

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

    NASA Astrophysics Data System (ADS)

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

    2013-09-01

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

  20. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail

    2010-12-01

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this

  1. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    SciTech Connect

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail,

    2010-12-23

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this

  2. A comparison of geospatially modeled fire behavior and potential application to fire and fuels management for the Savannah River Site.

    SciTech Connect

    Kurth, Laurie; Hollingsworth, LaWen; Shea, Dan

    2011-12-20

    This study evaluates modeled fire behavior for the Savannah River Site in the Atlantic Coastal Plain of the southeastern U.S. using three data sources: FCCS, LANDFIRE, and SWRA. The Fuel Characteristic Classification System (FCCS) was used to build fuelbeds from intensive field sampling of 629 plots. Custom fire behavior fuel models were derived from these fuelbeds. LANDFIRE developed surface fire behavior fuel models and canopy attributes for the U.S. using satellite imagery informed by field data. The Southern Wildfire Risk Assessment (SWRA) developed surface fire behavior fuel models and canopy cover for the southeastern U.S. using satellite imagery.

  3. Development and implementation of a finite element solution of the coupled neutron transport and thermoelastic equations governing the behavior of small nuclear assemblies

    NASA Astrophysics Data System (ADS)

    Wilson, Stephen Christian

    Small, highly enriched reactors designed for weapons effects simulations undergo extreme thermal transients during pulsed operations. The primary shutdown mechanism of these reactors---thermal expansion of fuel material---experiences an inertial delay resulting in a different value for the fuel temperature coefficient of reactivity during pulse operation as compared to the value appropriate for steady-state operation. The value appropriate for pulsed operation may further vary as a function of initial reactivity addition. Here we design and implement a finite element numerical method to predict the pulse operation behavior of Sandia Pulsed Reactor (SPR) II, SPR III, and a hypothetical spherical assembly with identical fuel properties without using operationally observed data in our model. These numerical results are compared to available SPR II and SPR III operational data. The numerical methods employed herein may be modified and expanded in functionality to provide both accurate characterization of the behavior of fast burst reactors of any common geometry or isotropic fuel material in the design phase, as well as a computational tool for general coupled thermomechanical-neutronics behavior in the solid state for any reactor type.

  4. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR... and equipment to extremely high standards is necessary in order to ensure predictable and safe...

  5. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR... and equipment to extremely high standards is necessary in order to ensure predictable and safe...

  6. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR... and equipment to extremely high standards is necessary in order to ensure predictable and safe...

  7. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR... and equipment to extremely high standards is necessary in order to ensure predictable and safe...

  8. Creep Behavior of Glass/Ceramic Sealant and its Effect on Long-term Performance of Solid Oxide Fuel Cells

    SciTech Connect

    Liu, Wenning N.; Sun, Xin; Koeppel, Brian J.; Stephens, Elizabeth V.; Khaleel, Mohammad A.

    2009-10-14

    The creep behavior of glass or glass-ceramic sealant materials used in solid oxide fuel cells (SOFCs) becomes relevant under SOFC operating temperatures. In this paper, the creep of glass-ceramic sealants was experimentally examined, and a standard linear solid model was applied to capture the creep behavior of glass ceramic sealant materials developed for planar SOFCs at high temperatures. The parameters of this model were determined based on the creep test results. Furthermore, the creep model was incorporated into finite-element software programs SOFC-MP and Mentat-FC developed at Pacific Northwest National Laboratory for multi-physics simulation of SOFCs. The effect of creep of glass ceramic sealant materials on the long-term performance of SOFC stacks was investigated by studying the stability of the flow channels and the stress redistribution in the glass seal and on the various interfaces of the glass seal with other layers. Finite element analyses were performed to quantify the stresses in various parts. The stresses in glass seals were released because of creep behavior during operations.

  9. Development of the thermal behavior analysis code DIRAD and the fuel design procedure for LMFBR

    NASA Astrophysics Data System (ADS)

    Nakae, N.; Tanaka, K.; Nakajima, H.; Matsumoto, M.

    1992-06-01

    It is very important to increase the fuel linear heat rating for improvement of economy in LMFBR without any degradation in safety. A reduction of the design margin is helpful to achieve the high power operation. The development of a fuel design code and a design procedure is effective on the reduction of the design margin. The thermal behavior analysis code DIRAD has been developed with respect to fuel restructuring and gap conductance models. These models have been calibrated and revised using irradiation data of fresh fuel. It is, therefore, found that the code is applicable for the thermal analysis with fresh fuel. The uncertainties in fuel irradiation condition and fuel fabrication tolerance together with the uncertainty of the code prediction have major contributions to the design margin. In the current fuel design the first two uncertainties independently contribute to temperature increment. Another method which can rationally explain the effect of the uncertainties on the temperature increment is adopted here. Then, the design margin may be rationally reduced.

  10. Analysis of Ignition Behavior in a Turbocharged Direct Injection Dual Fuel Engine Using Propane and Methane as Primary Fuels

    SciTech Connect

    Polk, A. C.; Gibson, C. M.; Shoemaker, N. T.; Srinivasan, K. K.; Krishnan, S. R.

    2013-05-24

    This paper presents experimental analyses of the ignition delay (ID) behavior for diesel-ignited propane and diesel-ignited methane dual fuel combustion. Two sets of experiments were performed at a constant speed (1800 rev/min) using a 4-cylinder direct injection diesel engine with the stock ECU and a wastegated turbocharger. First, the effects of fuel-air equivalence ratios (© pilot ¼ 0.2-0.6 and © overall ¼ 0.2-0.9) on IDs were quantified. Second, the effects of gaseous fuel percent energy substitution (PES) and brake mean effective pressure (BMEP) (from 2.5 to 10 bar) on IDs were investigated. With constant © pilot (> 0.5), increasing © overall with propane initially decreased ID but eventually led to premature propane autoignition; however, the corresponding effects with methane were relatively minor. Cyclic variations in the start of combustion (SOC) increased with increasing © overall (at constant © pilot), more significantly for propane than for methane. With increasing PES at constant BMEP, the ID showed a nonlinear (initially increasing and later decreasing) trend at low BMEPs for propane but a linearly decreasing trend at high BMEPs. For methane, increasing PES only increased IDs at all BMEPs. At low BMEPs, increasing PES led to significantly higher cyclic SOC variations and SOC advancement for both propane and methane. Finally, the engine ignition delay (EID) was also shown to be a useful metric to understand the influence of ID on dual fuel combustion.

  11. Failure Behavior Characterization of Mo-Modified Ti Surface by Impact Test and Finite Element Analysis

    NASA Astrophysics Data System (ADS)

    Ma, Yong; Qin, Jianfeng; Zhang, Xiangyu; Lin, Naiming; Huang, Xiaobo; Tang, Bin

    2015-07-01

    Using the impact test and finite element simulation, the failure behavior of the Mo-modified layer on pure Ti was investigated. In the impact test, four loads of 100, 300, 500, and 700 N and 104 impacts were adopted. The three-dimensional residual impact dents were examined using an optical microscope (Olympus-DSX500i), indicating that the impact resistance of the Ti surface was improved. Two failure modes cohesive and wearing were elucidated by electron backscatter diffraction and energy-dispersive spectrometer performed in a field-emission scanning electron microscope. Through finite element forward analysis performed at a typical impact load of 300 N, stress-strain distributions in the Mo-modified Ti were quantitatively determined. In addition, the failure behavior of the Mo-modified layer was determined and an ideal failure model was proposed for high-load impact, based on the experimental and finite element forward analysis results.

  12. Misconceptions concerning the behavior, fate and transport of the fuel oxygenates TBA and MTBE

    NASA Astrophysics Data System (ADS)

    Woodward, R.; Sloan, R.

    2003-04-01

    The release of gasoline from underground storage tanks and the subsequent appearance of dissolved constituents in drinking water has focused attention on the use of MTBE in reformulated fuels. Natural biodegradation of MTBE in soil, photo-oxidation in the atmosphere or chemical oxidation during remediation of gasoline releases can produce the intermediate tertiary butyl alcohol (TBA). TBA is also a fuel oxygenate and can be found as a co-product in MTBE synthesized from methanol and TBA. Because the physical properties of ethers and alcohols differ somewhat from the predominant hydrocarbon compounds in gasoline, misconceptions have developed about the behavior of fuel oxygenates in storage and in the subsurface. Critical review of several misconceptions about MTBE and TBA in gasoline reveals the concepts were conceived to rationalize early field observations and/or incomplete data sets. Closer scrutiny, in light of recent laboratory investigations, field data, case studies and world literature, clarifies these misconceptions and assumptions about the behavior of ether oxygenates and their degradation products in the environment. Commonly held misconceptions focus on four general areas of fuel and fuel oxygenate management: storage/dispensing, hydrology, remediation, and health effects. Storage/dispensing misconceptions address materials stability to ethers and alcohols in fuel and the environmental forensics of fuel systems failure. Groundwater and hydrology misconceptions deal with plume dynamics and the impact of fuel on drinking water resources. Remediation misconceptions focus on the performance of traditional hydrocarbon remediation technologies, recent developments in biodegradation and natural attenuation, drivers of remedial design and remediation costs. Health effects misconceptions address both acute and chronic exposure risk evaluations by national and international health agencies. Generally MTBE and TBA are manageable by the same processes and

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

    SciTech Connect

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

    2014-05-01

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

  14. Multifractal behavior of commodity markets: Fuel versus non-fuel products

    NASA Astrophysics Data System (ADS)

    Delbianco, Fernando; Tohmé, Fernando; Stosic, Tatijana; Stosic, Borko

    2016-09-01

    We investigate multifractal properties of commodity time series using multifractal detrended fluctuation analysis (MF-DFA). We find that agricultural and energy-related commodities exhibit very similar behavior, while the multifractal behavior of daily and monthly commodity series is rather different. Daily series demonstrate overall uncorrelated behavior, lower degree of multifractality and the dominance of small fluctuations. On the other hand, monthly commodity series show overall persistent behavior, higher degree of multifractality and the dominance of large fluctuations. After shuffling the series, we find that the multifractality is due to a broad probability density function for daily commodities series, while for monthly commodities series multifractality is caused by both a broad probability density function and long term correlations.

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

    SciTech Connect

    Chisholm, W.P.

    1995-06-01

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

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

    SciTech Connect

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

    1993-01-20

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

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

    ERIC Educational Resources Information Center

    Lloyd, William G.; Davenport, Derek A.

    1980-01-01

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

  18. Time-dependent mechanical behavior of proton exchange membrane fuel cell electrodes

    NASA Astrophysics Data System (ADS)

    Lu, Zongwen; Santare, Michael H.; Karlsson, Anette M.; Busby, F. Colin; Walsh, Peter

    2014-01-01

    The electrodes used for Proton Exchange Membrane Fuel Cells (PEMFCs) are typically painted or sprayed onto the membrane during manufacturing, making it difficult to directly characterize their mechanical behavior as a stand-alone material. An experimental-numerical hybrid technique is devised to extract the electrode properties from the experimentally measured properties of Nafion® 211 membrane

  19. Morphology-Induced Collective Behaviors: Dynamic Pattern Formation in Water-Floating Elements

    PubMed Central

    Nakajima, Kohei; Ngouabeu, Aubery Marchel Tientcheu; Miyashita, Shuhei; Göldi, Maurice; Füchslin, Rudolf Marcel; Pfeifer, Rolf

    2012-01-01

    Complex systems involving many interacting elements often organize into patterns. Two types of pattern formation can be distinguished, static and dynamic. Static pattern formation means that the resulting structure constitutes a thermodynamic equilibrium whose pattern formation can be understood in terms of the minimization of free energy, while dynamic pattern formation indicates that the system is permanently dissipating energy and not in equilibrium. In this paper, we report experimental results showing that the morphology of elements plays a significant role in dynamic pattern formation. We prepared three different shapes of elements (circles, squares, and triangles) floating in a water-filled container, in which each of the shapes has two types: active elements that were capable of self-agitation with vibration motors, and passive elements that were mere floating tiles. The system was purely decentralized: that is, elements interacted locally, and subsequently elicited global patterns in a process called self-organized segregation. We showed that, according to the morphology of the selected elements, a different type of segregation occurs. Also, we quantitatively characterized both the local interaction regime and the resulting global behavior for each type of segregation by means of information theoretic quantities, and showed the difference for each case in detail, while offering speculation on the mechanism causing this phenomenon. PMID:22715370

  20. Expected behavior of plutonium in the IFR fuel cycle

    NASA Astrophysics Data System (ADS)

    Steunenberg, R. K.; Johnson, I.

    The Integral Fast Reactor (IFR) is a metal-fueled, sodium-cooled reactor that will consist initially of a U-Zr alloy core in which the enriched uranium will be replaced gradually by plutonium bred in a uranium blanket. The plutonium is concentrated to the required level by extraction from the molten blanket material with a CaCl2-BaCl2 salt containing MgCl2 as an oxidant (halide slagging). The CaCl2-BaCl2 salt containing dissolved PuCl3 and UCl3 is added to the core process where fission products are removed by electrorefining, using a liquid cadmium anode, a metal cathode, and a LiCl-NaCl-CaCl2-BaCl2 molten salt electrolyte. The product is recovered as a metallic deposit on the cathode. The Halide slagging step is operated at about 1250 deg and the electrorefining step at about 450 C. These processes are expected to give low fission-product decontamination factors of the order of 100.

  1. The effect of oxygen as a light element in metallic liquids on partitioning behavior

    NASA Astrophysics Data System (ADS)

    Chabot, Nancy L.; Wollack, E. Alex; Humayun, Munir; Shank, Ellen M.

    2015-04-01

    Oxygen has been considered a potentially important light element in metallic liquids during a range of planetary processes, yet the influence of O in a metallic melt on element partitioning behavior is largely unknown. To investigate the effect of O in such systems, we conducted experiments in the Fe-S-O system, doped with 25 trace elements, which produced two immiscible metallic liquids. Our results indicate that the presence of O in the metallic liquid produces a distinctive chemical signature for W and Ga in particular. Tungsten shows an affinity for O in the metallic liquid and partitions more strongly into the metallic melt in the presence of O. The partitioning of Ga is relatively constant despite the presence of O, which is in contrast to the majority of the other siderophile elements in the study. Our experiments from 1400 to 1600 °C show no significant effect from temperature on the partitioning behavior of any trace elements over this limited temperature range. This distinctive chemical signature due to the presence of O in the metallic liquid has potential implications for modeling core formation, evaluating isotopic signatures produced by core crystallization, and interpreting chemical assemblages observed in meteorites.

  2. Limitation of Finite Element Analysis of Poroelastic Behavior of Biological Tissues Undergoing Rapid Loading

    PubMed Central

    Stokes, Ian A.; Chegini, Salman; Ferguson, Stephen J.; Gardner-Morse, Mack G.; Iatridis, James C.; Laible, Jeffrey P.

    2010-01-01

    The finite element method is used in biomechanics to provide numerical solutions to simulations of structures having complex geometry and spatially differing material properties. Time-varying load deformation behaviors can result from solid viscoelasticity as well as viscous fluid flow through porous materials. Finite element poroelastic analysis of rapidly loaded slow-draining materials may be ill-conditioned, but this problem is not widely known in the biomechanics field. It appears as instabilities in the calculation of interstitial fluid pressures, especially near boundaries and between different materials. Accurate solutions can require impractical compromises between mesh size and time steps. This article investigates the constraints imposed by this problem on tissues representative of the intervertebral disc, subjected to moderate physiological rates of deformation. Two test cylindrical structures were found to require over 104 linear displacement-constant pressure elements to avoid serious oscillations in calculated fluid pressure. Fewer Taylor–Hood (quadratic displacement–linear pressure elements) were required, but with complementary increases in computational costs. The Vermeer–Verruijt criterion for 1D mesh size provided guidelines for 3D mesh sizes for given time steps. Pressure instabilities may impose limitations on the use of the finite element method for simulating fluid transport behaviors of biological soft tissues at moderately rapid physiological loading rates. PMID:20306136

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

    NASA Astrophysics Data System (ADS)

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

    2009-07-01

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

  4. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    SciTech Connect

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed.

  5. Effects of controlled element dynamics on human feedforward behavior in ramp-tracking tasks.

    PubMed

    Laurense, Vincent A; Pool, Daan M; Damveld, Herman J; van Paassen, Marinus René M; Mulder, Max

    2015-02-01

    In real-life manual control tasks, human controllers are often required to follow a visible and predictable reference signal, enabling them to use feedforward control actions in conjunction with feedback actions that compensate for errors. Little is known about human control behavior in these situations. This paper investigates how humans adapt their feedforward control dynamics to the controlled element dynamics in a combined ramp-tracking and disturbance-rejection task. A human-in-the-loop experiment is performed with a pursuit display and vehicle-like controlled elements, ranging from a single integrator through second-order systems with a break frequency at either 3, 2, or 1 rad/s, to a double integrator. Because the potential benefits of feedforward control increase with steeper ramp segments in the target signal, three steepness levels are tested to investigate their possible effect on feedforward control with the various controlled elements. Analyses with four novel models of the operator, fitted to time-domain data, reveal feedforward control for all tested controlled elements and both (nonzero) tested levels of ramp steepness. For the range of controlled element dynamics investigated, it is found that humans adapt to these dynamics in their feedforward response, with a close to perfect inversion of the controlled element dynamics. No significant effects of ramp steepness on the feedforward model parameters are found.

  6. Understanding compressive deformation behavior of porous Ti using finite element analysis.

    PubMed

    Roy, Sandipan; Khutia, Niloy; Das, Debdulal; Das, Mitun; Balla, Vamsi Krishna; Bandyopadhyay, Amit; Chowdhury, Amit Roy

    2016-07-01

    In the present study, porous commercially pure (CP) Ti samples with different volume fraction of porosities were fabricated using a commercial additive manufacturing technique namely laser engineered net shaping (LENS™). Mechanical behavior of solid and porous samples was evaluated at room temperature under quasi-static compressive loading. Fracture surfaces of the failed samples were analyzed to determine the failure modes. Finite Element (FE) analysis using representative volume element (RVE) model and micro-computed tomography (CT) based model have been performed to understand the deformation behavior of laser deposited solid and porous CP-Ti samples. In vitro cell culture on laser processed porous CP-Ti surfaces showed normal cell proliferation with time, and confirmed non-toxic nature of these samples. PMID:27127074

  7. Simulation of the Dynamic Behavior of Electric Power Steering Systems Using Dedicated Finite Elements

    NASA Astrophysics Data System (ADS)

    Besson, François; Ferraris, Guy; Guingand, Michèle; Vaujany, Jean-Pierre De

    During the last decade, many new technical solutions dedicated to the comfort of automotive vehicle's drivers have raised, like Electrical Power Steering (EPS). To fulfill the more and more demanding requirements in terms of vibration and acoustics, the dynamic behavior of the whole steering is studied. The system is divided into dedicated finite elements (FE) describing the whole steering. The stress was first put on the gears models (worm gear and rack-and-pinion) and their anti-backlash systems as they have been identified as potential vibration sources. Mechanical non-linearities (clearances, non-linear stiffness) of the mechanical system are taken into account in these models. Then, this model allows simulating the transient response of the system to an input excitation. Each developed element is validated using a fitted experimental test bench. Then, the general model is correlated the same way. Hence models can be used to study the dynamic behavior of EPS systems or sub-systems.

  8. 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.

  9. Fuel moisture content enhances nonadditive effects of plant mixtures on flammability and fire behavior.

    PubMed

    Blauw, Luke G; Wensink, Niki; Bakker, Lisette; van Logtestijn, Richard S P; Aerts, Rien; Soudzilovskaia, Nadejda A; Cornelissen, J Hans C

    2015-09-01

    Fire behavior of plant mixtures includes a complex set of processes for which the interactive contributions of its drivers, such as plant identity and moisture, have not yet been unraveled fully. Plant flammability parameters of species mixtures can show substantial deviations of fire properties from those expected based on the component species when burnt alone; that is, there are nonadditive mixture effects. Here, we investigated how fuel moisture content affects nonadditive effects in fire behavior. We hypothesized that both the magnitude and variance of nonadditivity in flammability parameters are greater in moist than in dry fuel beds. We conducted a series of experimental burns in monocultures and 2-species mixtures with two ericaceous dwarf shrubs and two bryophyte species from temperate fire-prone heathlands. For a set of fire behavior parameters, we found that magnitude and variability of nonadditive effects are, on average, respectively 5.8 and 1.8 times larger in moist (30% MC) species mixtures compared to dry (10% MC) mixed fuel beds. In general, the moist mixtures caused negative nonadditive effects, but due to the larger variability these mixtures occasionally caused large positive nonadditive effects, while this did not occur in dry mixtures. Thus, at moister conditions, mixtures occasionally pass the moisture threshold for ignition and fire spread, which the monospecific fuel beds are unable to pass. We also show that the magnitude of nonadditivity is highly species dependent. Thus, contrary to common belief, the strong nonadditive effects in mixtures can cause higher fire occurrence at moister conditions. This new integration of surface fuel moisture and species interactions will help us to better understand fire behavior in the complexity of natural ecosystems. PMID:26380709

  10. Fuel moisture content enhances nonadditive effects of plant mixtures on flammability and fire behavior

    PubMed Central

    Blauw, Luke G; Wensink, Niki; Bakker, Lisette; van Logtestijn, Richard S P; Aerts, Rien; Soudzilovskaia, Nadejda A; Cornelissen, J Hans C

    2015-01-01

    Fire behavior of plant mixtures includes a complex set of processes for which the interactive contributions of its drivers, such as plant identity and moisture, have not yet been unraveled fully. Plant flammability parameters of species mixtures can show substantial deviations of fire properties from those expected based on the component species when burnt alone; that is, there are nonadditive mixture effects. Here, we investigated how fuel moisture content affects nonadditive effects in fire behavior. We hypothesized that both the magnitude and variance of nonadditivity in flammability parameters are greater in moist than in dry fuel beds. We conducted a series of experimental burns in monocultures and 2-species mixtures with two ericaceous dwarf shrubs and two bryophyte species from temperate fire-prone heathlands. For a set of fire behavior parameters, we found that magnitude and variability of nonadditive effects are, on average, respectively 5.8 and 1.8 times larger in moist (30% MC) species mixtures compared to dry (10% MC) mixed fuel beds. In general, the moist mixtures caused negative nonadditive effects, but due to the larger variability these mixtures occasionally caused large positive nonadditive effects, while this did not occur in dry mixtures. Thus, at moister conditions, mixtures occasionally pass the moisture threshold for ignition and fire spread, which the monospecific fuel beds are unable to pass. We also show that the magnitude of nonadditivity is highly species dependent. Thus, contrary to common belief, the strong nonadditive effects in mixtures can cause higher fire occurrence at moister conditions. This new integration of surface fuel moisture and species interactions will help us to better understand fire behavior in the complexity of natural ecosystems. PMID:26380709

  11. Fuel moisture content enhances nonadditive effects of plant mixtures on flammability and fire behavior.

    PubMed

    Blauw, Luke G; Wensink, Niki; Bakker, Lisette; van Logtestijn, Richard S P; Aerts, Rien; Soudzilovskaia, Nadejda A; Cornelissen, J Hans C

    2015-09-01

    Fire behavior of plant mixtures includes a complex set of processes for which the interactive contributions of its drivers, such as plant identity and moisture, have not yet been unraveled fully. Plant flammability parameters of species mixtures can show substantial deviations of fire properties from those expected based on the component species when burnt alone; that is, there are nonadditive mixture effects. Here, we investigated how fuel moisture content affects nonadditive effects in fire behavior. We hypothesized that both the magnitude and variance of nonadditivity in flammability parameters are greater in moist than in dry fuel beds. We conducted a series of experimental burns in monocultures and 2-species mixtures with two ericaceous dwarf shrubs and two bryophyte species from temperate fire-prone heathlands. For a set of fire behavior parameters, we found that magnitude and variability of nonadditive effects are, on average, respectively 5.8 and 1.8 times larger in moist (30% MC) species mixtures compared to dry (10% MC) mixed fuel beds. In general, the moist mixtures caused negative nonadditive effects, but due to the larger variability these mixtures occasionally caused large positive nonadditive effects, while this did not occur in dry mixtures. Thus, at moister conditions, mixtures occasionally pass the moisture threshold for ignition and fire spread, which the monospecific fuel beds are unable to pass. We also show that the magnitude of nonadditivity is highly species dependent. Thus, contrary to common belief, the strong nonadditive effects in mixtures can cause higher fire occurrence at moister conditions. This new integration of surface fuel moisture and species interactions will help us to better understand fire behavior in the complexity of natural ecosystems.

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

    NASA Technical Reports Server (NTRS)

    Walton, James T.

    1992-01-01

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

  13. The wind-evoked escape behavior of the cricket Gryllus bimaculatus: integration of behavioral elements

    PubMed

    Tauber; Camhi

    1995-01-01

    The wind-evoked escape behavior of freely ranging crickets (Gryllus bimaculatus) was studied using high-speed video and film analysis. The escape response can be of three types: a turn, a jump or a turn + jump. Any of these can be followed by running. The turn is similar to that of the cockroach, in terms of the details of body and leg movements. A jump occurs only when the cricket has its back to the wind, either because the stimulus came approximately from behind or because the cricket had first turned away from the wind and then jumped. The jump, like that of locust, requires some form of energy storage and quick release to obtain the necessary power. Locusts use long-term co-activation of antagonistic leg motor neurons to produce mechanical energy storage. By contrast, crickets do not appear to co-activate antagonistic leg motor neurons. Possible alternative energy storage and release mechanisms are discussed.

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

    SciTech Connect

    Edlund, D.J.

    1996-12-31

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

  15. Coupled thermomechanical behavior of graphene using the spring-based finite element approach

    NASA Astrophysics Data System (ADS)

    Georgantzinos, S. K.; Giannopoulos, G. I.; Anifantis, N. K.

    2016-07-01

    The prediction of the thermomechanical behavior of graphene using a new coupled thermomechanical spring-based finite element approach is the aim of this work. Graphene sheets are modeled in nanoscale according to their atomistic structure. Based on molecular theory, the potential energy is defined as a function of temperature, describing the interatomic interactions in different temperature environments. The force field is approached by suitable straight spring finite elements. Springs simulate the interatomic interactions and interconnect nodes located at the atomic positions. Their stiffness matrix is expressed as a function of temperature. By using appropriate boundary conditions, various different graphene configurations are analyzed and their thermo-mechanical response is approached using conventional finite element procedures. A complete parametric study with respect to the geometric characteristics of graphene is performed, and the temperature dependency of the elastic material properties is finally predicted. Comparisons with available published works found in the literature demonstrate the accuracy of the proposed method.

  16. Combustion and leaching behavior of elements in the argonne premium coal samples

    USGS Publications Warehouse

    Finkelman, R.B.; Palmer, C.A.; Krasnow, M.R.; Aruscavage, P. J.; Sellers, G.A.; Dulong, F.T.

    1990-01-01

    Eight Argonne Premium Coal samples and two other coal samples were used to observe the effects of combustion and leaching on 30 elements. The results were used to infer the modes of occurrence of these elements. Instrumental neutron activation analysis indicates that the effects of combustion and leaching on many elements varied markedly among the samples. As much as 90% of the selenium and bromine is volatilized from the bituminous coal samples, but substantially less is volatilized from the low-rank coals. We interpret the combustion and leaching behavior of these elements to indicate that they are associated with the organic fraction. Sodium, although nonvolatile, is ion-exchangeable in most samples, particularly in the low-rank coal samples where it is likely to be associated with the organic constituents. Potassium is primarily in an ion-exchangeable form in the Wypdak coal but is in HF-soluble phases (probably silicates) in most other samples. Cesium is in an unidentified HNO3-soluble phase in most samples. Virtually all the strontium and barium in the low-rank coal samples is removed by NH4OAc followed by HCl, indicating that these elements probably occur in both organic and inorganic phases. Most tungsten and tantalum are in insoluble phases, perhaps as oxides or in organic association. Hafnium is generally insoluble, but as much as 65% is HF soluble, perhaps due to the presence of very fine grained or metamict zircon. We interpret the leaching behavior of uranium to indicate its occurrence in chelates and its association with silicates and with zircon. Most of the rare-earth elements (REE) and thorium appear to be associated with phosphates. Differences in textural relationships may account for some of the differences in leaching behavior of the REE among samples. Zinc occurs predominantly in sphalerite. Either the remaining elements occur in several different modes of occurrence (scandium, iron), or the leaching data are equivocal (arsenic, antimony

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

    SciTech Connect

    Willingham, C.E.

    1981-10-01

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

  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. Effects of nanoparticle zinc oxide on emotional behavior and trace elements homeostasis in rat brain.

    PubMed

    Amara, Salem; Slama, Imen Ben; Omri, Karim; El Ghoul, Jaber; El Mir, Lassaad; Rhouma, Khemais Ben; Abdelmelek, Hafedh; Sakly, Mohsen

    2015-12-01

    Over recent years, nanotoxicology and the potential effects on human body have grown in significance, the potential influences of nanosized materials on the central nervous system have received more attention. The aim of this study was to determine whether zinc oxide (ZnO) nanoparticles (NPs) exposure cause alterations in emotional behavior and trace elements homeostasis in rat brain. Rats were treated by intraperitoneal injection of ZnO NPs (20-30 nm) at a dose of 25 mg/kg body weight. Sub -: acute ZnO NPs treatment induced no significant increase in the zinc content in the homogenate brain. Statistically significant decreases in iron and calcium concentrations were found in rat brain tissue compared to control. However, sodium and potassium contents remained unchanged. Also, there were no significant changes in the body weight and the coefficient of brain. In the present study, the anxiety-related behavior was evaluated using the plus-maze test. ZnO NPs treatment modulates slightly the exploratory behaviors of rats. However, no significant differences were observed in the anxious index between ZnO NP-treated rats and the control group (p > 0.05). Interestingly, our results demonstrated minimal effects of ZnO NPs on emotional behavior of animals, but there was a possible alteration in trace elements homeostasis in rat brain.

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

    NASA Technical Reports Server (NTRS)

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

    2016-01-01

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

  1. Finite element analysis of the dynamic behavior of a laminated windscreen with frequency dependent viscoelastic core.

    PubMed

    Bouayed, Kaïss; Hamdi, Mohamed-Ali

    2012-08-01

    This paper presents numerical and experimental validation of results obtained by a shell finite element, which has been developed for modeling of the dynamic behavior of sandwich multilayered structures with a viscoelastic core. The proposed shell finite element is very easy to implement in existing finite element solvers, since it uses only the displacements as degrees of freedom at external faces and at inter-layer interfaces. The displacement field is linearly interpolated in the thickness direction of each layer, and analytical integration is made in the thickness direction in order to avoid meshing of each sandwich layer by solid elements. Only the two dimensional mid-surface of reference is meshed, facilitating the mesh generation task. A simplified modal approach using a real modal basis is also proposed to efficiently calculate the dynamic response of the sandwich structure. The proposed method reduces the memory size and computing time and takes into account the frequency-dependence of the polymer core mechanical properties. Results obtained by the proposed element in conjunction with the simplified modal method have been numerically and experimentally validated by comparison to results obtained by commercial software codes (MSC/NASTRAN and ESI/RAYON-VTM), and to measurements done on automobile windscreens. PMID:22894198

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

    SciTech Connect

    Pitner, A.L.

    1998-08-21

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

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

  4. Mathematical modeling of minor-element behavior in flash smelting of copper concentrates and flash converting of copper mattes

    NASA Astrophysics Data System (ADS)

    Chaubal, P. C.; Sohn, H. Y.; George, D. B.; Bailey, L. K.

    1989-02-01

    A mathematical model has been developed to describe the behavior of minor elements during flash smelting and flash converting. The model incorporates equations describing volatilization of minor elements from the molten particles and distribution of these elements between the molten phases in the settler. The basic premise of the volatilization model is that at the surface of the molten particle, the partial pressures of the minor-element species are those at equilibrium. Transport of the minor-element species to the gas then is described by external mass transfer. Good agreement has been obtained between observed and predicted behaviors. The effects of oxygen enrichment, matte grade, and wall temperature, as well as the bath temperature, on minor-element behavior have been elucidated.

  5. Mechanical Behavior of Free-Standing Fuel Cell Electrodes on Water Surface.

    PubMed

    Kim, Sanwi; Kim, Jae-Han; Oh, Jong-Gil; Jang, Kyung-Lim; Jeong, Byeong-Heon; Hong, Bo Ki; Kim, Taek-Soo

    2016-06-22

    Fundamental understanding of the mechanical behavior of polymer electrolyte fuel cell electrodes as free-standing materials is essential to develop mechanically robust fuel cells. However, this has been a significant challenge due to critical difficulties, such as separating the pristine electrode from the substrate without damage and precisely measuring the mechanical properties of the very fragile and thin electrodes. We report the mechanical behavior of free-standing fuel cell electrodes on the water surface through adopting an innovative ice-assisted separation method to separate the electrode from decal transfer film. It is found that doubling the ionomer content in electrodes increases not only the tensile stress at the break and the Young's modulus (E) of the electrodes by approximately 2.1-3.5 and 1.7-2.4 times, respectively, but also the elongation at the break by approximately 1.5-1.7 times, which indicates that stronger, stiffer, and tougher electrodes are attained with increasing ionomer content, which have been of significant interest in materials research fields. The scaling law relationship between Young's modulus and density (ρ) has been unveiled as E ∼ ρ(1.6), and it is compared with other materials. These findings can be used to develop mechanically robust electrodes for fuel cell applications.

  6. Mechanical Behavior of Free-Standing Fuel Cell Electrodes on Water Surface.

    PubMed

    Kim, Sanwi; Kim, Jae-Han; Oh, Jong-Gil; Jang, Kyung-Lim; Jeong, Byeong-Heon; Hong, Bo Ki; Kim, Taek-Soo

    2016-06-22

    Fundamental understanding of the mechanical behavior of polymer electrolyte fuel cell electrodes as free-standing materials is essential to develop mechanically robust fuel cells. However, this has been a significant challenge due to critical difficulties, such as separating the pristine electrode from the substrate without damage and precisely measuring the mechanical properties of the very fragile and thin electrodes. We report the mechanical behavior of free-standing fuel cell electrodes on the water surface through adopting an innovative ice-assisted separation method to separate the electrode from decal transfer film. It is found that doubling the ionomer content in electrodes increases not only the tensile stress at the break and the Young's modulus (E) of the electrodes by approximately 2.1-3.5 and 1.7-2.4 times, respectively, but also the elongation at the break by approximately 1.5-1.7 times, which indicates that stronger, stiffer, and tougher electrodes are attained with increasing ionomer content, which have been of significant interest in materials research fields. The scaling law relationship between Young's modulus and density (ρ) has been unveiled as E ∼ ρ(1.6), and it is compared with other materials. These findings can be used to develop mechanically robust electrodes for fuel cell applications. PMID:27183314

  7. Behavior of major and trace elements during weathering of sericite-quartz schist

    NASA Astrophysics Data System (ADS)

    Gong, Qingjie; Deng, Jun; Yang, Liqiang; Zhang, Jing; Wang, Qingfei; Zhang, Gaixia

    2011-07-01

    Two regolith profiles developed on the sericite-quartz schist in subtropical humid environment were selected to investigate behaviors of major and trace elements during weathering in Mengman gold deposit of Yunnan province, China. One profile located in the mining district sheared by a fault and the other was outside the mining area which represented the normal weathering profile on the schist. Regolith samples were collected in both profiles sequentially. Thirteen major oxides and 23 trace elements (including REE) were analyzed and their behaviors were compared in these two profiles. Based on the idea that immobile element is just a relative notion, we presented a method of immobile plateau to determine immobile elements during each stage in a progressive geochemical process and used mass ratio ( MR) to calculate the percentage of gain or loss ( X gp) of each element during the whole process. In both profiles, only TiO 2 was immobile during the whole weathering. The regolith profile formed on the mineralized schist recorded the weathering process more sensitively than the regolith profile on the normal schist. REE was mobile and fractionated during the schist weathering. LREE was loss in mass during the soil development stage which resulted from the chemical leaching, but was gain in mass during the pedogenesis stage because of the preferential absorption of soil to LREE. The LREE depletion near the fault during weathering was the collective effects of chemical leaching and physical accumulation. HFSE were all mobile in the mineralized regolith profile especially near the fault. But Nb-Ta and Zr-Hf were covariant in both profiles during the schist weathering.

  8. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    DOE PAGES

    Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; Novascone, Stephen R.; Perez, Danielle M.; Spencer, Benjamin W.; Luzzi, Lelio; Uffelen, Paul Van; Williamson, Richard L.

    2014-10-12

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code and a recently implemented physics-based model for the coupled fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information from the open literature. The study leads to an initial quantitative assessment of the uncertaintymore » in fission gas behavior modeling with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.« less

  9. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    NASA Astrophysics Data System (ADS)

    Pastore, Giovanni; Swiler, L. P.; Hales, J. D.; Novascone, S. R.; Perez, D. M.; Spencer, B. W.; Luzzi, L.; Van Uffelen, P.; Williamson, R. L.

    2015-01-01

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code with a recently implemented physics-based model for fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information in the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior predictions with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, significantly higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

  10. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    SciTech Connect

    Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; Novascone, Stephen R.; Perez, Danielle M.; Spencer, Benjamin W.; Luzzi, Lelio; Uffelen, Paul Van; Williamson, Richard L.

    2014-10-12

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code and a recently implemented physics-based model for the coupled fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information from the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior modeling with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

  11. Ultra-thin 242mAm fuel elements in nuclear reactors. II

    NASA Astrophysics Data System (ADS)

    Ronen, Y.; Raitses, G.

    2004-04-01

    There is growing interest in using 242mAm as a nuclear fuel for space reactors and nuclear batteries. In this paper, we discuss different 242mAm enrichments, as well as fuel weight requirements, to produce a critical reactor. It was found that relatively low enrichments of 242mAm, about 10 w/o, are enough to guarantee criticality. Such low enrichments might eliminate the need for a 242mAm enrichment process. It was also found that the best results for low 242mAm requirements are obtained with a moderator to fuel volume ratio of 10,000.

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

    PubMed

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

    2016-01-01

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

  13. Using fine-scale fuel measurements to assess wildland fuels, potential fire behavior and hazard mitigation treatments in the southeastern USA.

    SciTech Connect

    Ottmar, Roger, D.; Blake, John, I.; Crolly, William, T.

    2012-01-01

    The inherent spatial and temporal heterogeneity of fuelbeds in forests of the southeastern United States may require fine scale fuel measurements for providing reliable fire hazard and fuel treatment effectiveness estimates. In a series of five papers, an intensive, fine scale fuel inventory from the Savanna River Site in the southeastern United States is used for building fuelbeds and mapping fire behavior potential, evaluating fuel treatment options for effectiveness, and providing a comparative analysis of landscape modeled fire behavior using three different data sources including the Fuel Characteristic Classification System, LANDFIRE, and the Southern Wildfire Risk Assessment. The research demonstrates that fine scale fuel measurements associated with fuel inventories repeated over time can be used to assess broad scale wildland fire potential and hazard mitigation treatment effectiveness in the southeastern USA and similar fire prone regions. Additional investigations will be needed to modify and improve these processes and capture the true potential of these fine scale data sets for fire and fuel management planning.

  14. Finite element modeling of electromechanical behavior of a dielectric electroactive polymer actuator

    NASA Astrophysics Data System (ADS)

    Deodhar, Aseem; York, Alexander; Hodgins, Micah; Seelecke, Stefan

    2011-04-01

    Dielectric Electroactive Polymers (DEAP) will undergo large deformations when subject to an electric field making them an attractive material for use in novel actuator systems. There are many challenges with successful application and design of DEAP actuators resulting from their inherent electromechanical coupling and non-linear material behavior. FE modeling of the material behavior is a useful tool to better understand such systems and aid in the optimal design of prototypes. These modeling efforts must account for the electromechanical coupling in order to accurately predict their response to multiple loading conditions expected during real operating conditions. This paper presents a Finite Element model of a dielectric elastomer undergoing out-of-plane, axisymmetric deformation. The response of the elastomer was investigated while it was subjected to mechanical and electric fields and combined electro-mechanical actuation. The compliant electrodes have a large effect on the mechanical behavior of the EAP which needs to be taken into consideration while modeling the EAP as a system. The model is adapted to include the effect of electrode stiffness on the mechanical response of the actuator. The model was developed using the commercial Finite Element Modeling software, COMSOL. The results from the mechanical simulations are presented in the form of forcedisplacement curves and are validated with comparisons to experimental results. Electromechanical simulations are carried out and the stroke of the actuator for different electrode stiffness values is compared with experimental values when the EAP is biased with a constant force.

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

    SciTech Connect

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

    1997-12-01

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

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

    DOEpatents

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

    1999-01-01

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

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

    DOEpatents

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

    1999-03-23

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

  18. Corrosion behavior of dental alloys used for retention elements in prosthodontics.

    PubMed

    Nierlich, Judith; Papageorgiou, Spyridon N; Bourauel, Christoph; Hültenschmidt, Robert; Bayer, Stefan; Stark, Helmut; Keilig, Ludger

    2016-06-01

    The purpose of this study was to investigate the corrosion behavior of 10 different high noble gold-based dental alloys, used for prosthodontic retention elements, according to ISO 10271. Samples of 10 high-noble and noble gold-based dental alloys were subjected to: (i) static immersion tests with subsequent analysis of ion release for eight different elements using mass spectrometry; (ii) electrochemical tests, including open-circuit potential and potentiodynamic scans; and (iii) scanning electron microscopy, followed by energy-dispersive X-ray microscopy. The results were analyzed using one-way ANOVA and Sidak multiple-comparisons post-hoc test at a level of significance of α = 0.05. Significant differences were found among the 10 alloys studied for all ions (P < 0.001). The potentiodynamic analysis showed values from -82.5 to 102.8 mV for the open-circuit potential and from 566.7 to 1367.5 mV for the breakdown potential. Both the open-circuit and the breakdown potential varied considerably among these alloys. Scanning electron microscopy analysis confirmed the existence of typically small-diameter corrosion defects, whilst the energy-dispersive X-ray analysis found no significant alteration in the elemental composition of the alloys. The results of this study reveal the variability in the corrosive resistance among the materials used for retention elements in prosthodontics. PMID:27061513

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

    NASA Astrophysics Data System (ADS)

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

    2012-12-01

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

  20. Simulation of micromechanical behavior of polycrystals: finite elements vs. fast Fourier transforms

    SciTech Connect

    Lebensohn, Ricardo A; Prakash, Arun

    2009-01-01

    In this work, we compare finite element and fast Fourier transform approaches for the prediction of micromechanical behavior of polycrystals. Both approaches are full-field approaches and use the same visco-plastic single crystal constitutive law. We investigate the texture and the heterogeneity of the inter- and intragranular, stress and strain fields obtained from the two models. Additionally, we also look into their computational performance. Two cases - rolling of aluminium and wire drawing of tungsten - are used to evaluate the predictions of the two mode1s. Results from both the models are similar, when large grain distortions do not occur in the polycrystal. The finite element simulations were found to be highly computationally intensive, in comparison to the fast Fourier transform simulations.

  1. Finite element modeling as a tool for predicting the fracture behavior of robocast scaffolds.

    PubMed

    Miranda, Pedro; Pajares, Antonia; Guiberteau, Fernando

    2008-11-01

    The use of finite element modeling to calculate the stress fields in complex scaffold structures and thus predict their mechanical behavior during service (e.g., as load-bearing bone implants) is evaluated. The method is applied to identifying the fracture modes and estimating the strength of robocast hydroxyapatite and beta-tricalcium phosphate scaffolds, consisting of a three-dimensional lattice of interpenetrating rods. The calculations are performed for three testing configurations: compression, tension and shear. Different testing orientations relative to the calcium phosphate rods are considered for each configuration. The predictions for the compressive configurations are compared to experimental data from uniaxial compression tests.

  2. Suspicious Behavior Detection System for an Open Space Parking Based on Recognition of Human Elemental Actions

    NASA Astrophysics Data System (ADS)

    Inomata, Teppei; Kimura, Kouji; Hagiwara, Masafumi

    Studies for video surveillance applications for preventing various crimes such as stealing and violence have become a hot topic. This paper proposes a new video surveillance system that can detect suspicious behaviors such as a car break-in and vandalization in an open space parking, and that is based on image processing. The proposed system has the following features: it 1)deals time series data flow, 2)recognizes “human elemental actions” using statistic features, and 3)detects suspicious behavior using Subspace method and AdaBoost. We conducted the experiments to test the performance of the proposed system using open space parking scenes. As a result, we obtained about 10.0% for false positive rate, and about 4.6% for false negative rate.

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

    SciTech Connect

    Pesic, Milan P

    2003-10-15

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

  4. The dynamic and steady state behavior of a PEM fuel cell as an electric energy source

    NASA Astrophysics Data System (ADS)

    Costa, R. A.; Camacho, J. R.

    The main objective of this work is to extract information on the internal behavior of three small polymer electrolyte membrane fuel cells under static and dynamic load conditions. A computational model was developed using Scilab [SCILAB 4, Scilab-a free scientific software package, http://www.scilab.org/, INRIA, France, December, 2005] to simulate the static and dynamic performance [J.M. Correa, A.F. Farret, L.N. Canha, An analysis of the dynamic performance of proton exchange membrane fuel cells using an electrochemical model, in: 27th Annual Conference of IEEE Industrial Electronics Society, 2001, pp. 141-146] of this particular type of fuel cell. This dynamic model is based on electrochemical equations and takes into consideration most of the chemical and physical characteristics of the device in order to generate electric power. The model takes into consideration the operating, design parameters and physical material properties. The results show the internal losses and concentration effects behavior, which are of interest for power engineers and researchers.

  5. Taxonomy for Strengthening the Identification of Core Elements for Evidence-Based Behavioral Interventions for HIV/AIDS Prevention

    ERIC Educational Resources Information Center

    Galbraith, Jennifer S.; Herbst, Jeffrey H.; Whittier, David K.; Jones, Patricia L.; Smith, Bryce D.; Uhl, Gary; Fisher, Holly H.

    2011-01-01

    The concept of core elements was developed to denote characteristics of an intervention, such as activities or delivery methods, presumed to be responsible for the efficacy of evidence-based behavioral interventions (EBIs) for HIV/AIDS prevention. This paper describes the development of a taxonomy of core elements based on a literature review of…

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

    NASA Technical Reports Server (NTRS)

    Springborn, R. H.

    1971-01-01

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

  7. Technology requirements for an orbiting fuel depot: A necessary element of a space infrastructure

    NASA Technical Reports Server (NTRS)

    Stubbs, R. M.; Corban, R. R.; Willoughby, A. J.

    1988-01-01

    Advanced planning within NASA has identified several bold space exploration initiatives. The successful implementation of these missions will require a supporting space infrastructure which would include a fuel depot, an orbiting facility to store, transfer and process large quantities of cryogenic fluids. In order to adequately plan the technology development programs required to enable the construction and operation of a fuel depot, a multidisciplinary workshop was convened to assess critical technologies and their state of maturity. Since technology requirements depend strongly on the depot design assumptions, several depot concepts are presented with their effect on criticality ratings. Over 70 depot-related technology areas are addressed.

  8. Technology requirements for an orbiting fuel depot - A necessary element of a space infrastructure

    NASA Technical Reports Server (NTRS)

    Stubbs, R. M.; Corban, R. R.; Willoughby, A. J.

    1988-01-01

    Advanced planning within NASA has identified several bold space exploration initiatives. The successful implementation of these missions will require a supporting space infrastructure which would include a fuel depot, an orbiting facility to store, transfer and process large quantities of cryogenic fluids. In order to adequately plan the technology development programs required to enable the construction and operation of a fuel depot, a multidisciplinary workshop was convened to assess critical technologies and their state of maturity. Since technology requirements depend strongly on the depot design assumptions, several depot concepts are presented with their effect of criticality ratings. Over 70 depot-related technology areas are addressed.

  9. Mechanical behavior of linear amorphous polymers: Comparison between molecular dynamics and finite-element simulations

    NASA Astrophysics Data System (ADS)

    Solar, Mathieu; Meyer, Hendrik; Gauthier, Christian; Fond, Christophe; Benzerara, Olivier; Schirrer, Robert; Baschnagel, Jörg

    2012-02-01

    This paper studies the rheology of weakly entangled polymer melts and films in the glassy domain and near the rubbery domain using two different methods: molecular dynamics (MD) and finite element (FE) simulations. In a first step, the uniaxial mechanical behavior of a bulk polymer sample is studied by means of particle-based MD simulations. The results are in good agreement with experimental data, and mechanical properties may be computed from the simulations. This uniaxial mechanical behavior is then implemented in FE simulations using an elasto-viscoelasto-viscoplastic constitutive law in a continuum mechanics (CM) approach. In a second step, the mechanical response of a polymer film during an indentation test is modeled with the MD method and with the FE simulations using the same constitutive law. Good agreement is found between the MD and CM results. This work provides evidence in favor of using MD simulations to investigate the local physics of contact mechanics, since the volume elements studied are representative and thus contain enough information about the microstructure of the polymer model, while surface phenomena (adhesion and surface tension) are naturally included in the MD approach.

  10. Mechanical behavior of linear amorphous polymers: comparison between molecular dynamics and finite-element simulations.

    PubMed

    Solar, Mathieu; Meyer, Hendrik; Gauthier, Christian; Fond, Christophe; Benzerara, Olivier; Schirrer, Robert; Baschnagel, Jörg

    2012-02-01

    This paper studies the rheology of weakly entangled polymer melts and films in the glassy domain and near the rubbery domain using two different methods: molecular dynamics (MD) and finite element (FE) simulations. In a first step, the uniaxial mechanical behavior of a bulk polymer sample is studied by means of particle-based MD simulations. The results are in good agreement with experimental data, and mechanical properties may be computed from the simulations. This uniaxial mechanical behavior is then implemented in FE simulations using an elasto-viscoelasto-viscoplastic constitutive law in a continuum mechanics (CM) approach. In a second step, the mechanical response of a polymer film during an indentation test is modeled with the MD method and with the FE simulations using the same constitutive law. Good agreement is found between the MD and CM results. This work provides evidence in favor of using MD simulations to investigate the local physics of contact mechanics, since the volume elements studied are representative and thus contain enough information about the microstructure of the polymer model, while surface phenomena (adhesion and surface tension) are naturally included in the MD approach. PMID:22463237

  11. Intervertebral disc creep behavior assessment through an open source finite element solver.

    PubMed

    Castro, A P G; Wilson, W; Huyghe, J M; Ito, K; Alves, J L

    2014-01-01

    Degenerative Disc Disease (DDD) is one of the largest health problems faced worldwide, based on lost working time and associated costs. By means of this motivation, this work aims to evaluate a biomimetic Finite Element (FE) model of the Intervertebral Disc (IVD). Recent studies have emphasized the importance of an accurate biomechanical modeling of the IVD, as it is a highly complex multiphasic medium. Poroelastic models of the disc are mostly implemented in commercial finite element packages with limited access to the algorithms. Therefore, a novel poroelastic formulation implemented on a home-developed open source FE solver is briefly addressed throughout this paper. The combination of this formulation with biphasic osmotic swelling behavior is also taken into account. Numerical simulations were devoted to the analysis of the non-degenerated human lumbar IVD time-dependent behavior. The results of the tests performed for creep assessment were inside the scope of the experimental data, with a remarkable improvement of the numerical accuracy when compared with previously published results obtained with ABAQUS(®). In brief, this in-development open-source FE solver was validated with literature experimental data and aims to be a valuable tool to study the IVD biomechanics and DDD mechanisms. PMID:24210477

  12. Trace element behavior in hydrothermal experiments: Implications for fluid processes at shallow depths in subduction zones

    NASA Astrophysics Data System (ADS)

    You, C.-F.; Castillo, P. R.; Gieskes, J. M.; Chan, L. H.; Spivack, A. J.

    1996-05-01

    Chemical evaluation of fluids affected during progressive water-sediment interactions provides critical information regarding the role of slab dehydration and/or crustal recycling in subduction zones. To place some constraints on geochemical processes during sediment subduction, reactions between décollement sediments and synthetic NaCl-CaCl 2 solutions at 25-350°C and 800 bar were monitored in laboratory hydrothermal experiments using an autoclave apparatus. This is the first attempt in a single set of experiments to investigate the relative mobilities of many subduction zone volatiles and trace elements but, because of difficulties in conducting hydrothermal experiments on sediments at high P-T conditions, the experiments could only be designed for a shallow (˜ 10 km) depth. The experimental results demonstrate mobilization of volatiles (B and NH 4) and incompatible elements (As, Be, Cs, Li, Pb, Rb) in hydrothermal fluids at relatively low temperatures (˜ 300°C). In addition, a limited fractionation of light from heavy rare earth elements (REEs) occurs under hydrothermal conditions. On the other hand, the high field strength elements (HFSEs) Cr, Hf, Nb, Ta, Ti, and Zr are not mobile in the reacted fluids. The observed behavior of volatiles and trace elements in hydrothermal fluids is similar to the observed enrichment in As, B, Cs, Li, Pb, Rb, and light REEs and depletion in HFSEs in arc magmas relative to magmas derived directly from the upper mantle. Thus, our work suggests a link between relative mobilities of trace elements in hydrothermal fluids and deep arc magma generation in subduction zones. The experimental results are highly consistent with the proposal that the addition of subduction zone hydrous fluids to the subarc mantle, which has been depleted by previous melting events, can produce the unique characteristics of arc magmas. Moreover, the results suggest that deeply subducted sediments may no longer have the composition necessary to generate

  13. Investigation on "saw-tooth" behavior of PEM fuel cell performance during shutdown and restart cycles

    NASA Astrophysics Data System (ADS)

    Qi, Zhigang; Tang, Hao; Guo, Qunhui; Du, Bin

    It was sometimes observed that the performance of a proton-exchange membrane fuel cell improved after the cell went through shutdown and restart cycles. Such a performance recovery led to a "saw-tooth" performance pattern when multiple shutdowns and restarts occurred during the endurance test of a fuel cell. The shutdowns included both planned shutdowns and unintended ones due to station trips or emergency stops (E-stops). The length of the shutdown periods ranged from a few minutes to several weeks. Although such a "saw-tooth" behavior could be attributed to multiple reasons such as: (1) catalyst surface oxidation state change; (2) catalyst surface cleansing; or (3) water management, we found that it was mainly related to water management in our cases after a systematic investigation employing both single cells and stacks.

  14. Trace elements in the Mississippi River Delta outflow region: Behavior at high discharge

    NASA Astrophysics Data System (ADS)

    Shiller, Alan M.; Boyle, Edward A.

    1991-11-01

    Samples for dissolved trace element analysis were collected in surface waters of the plume of the Mississippi River during a period of high river discharge. These field data are compared with results of laboratory mixing experiments. The studies show that Cu, Ni, and Mo are largely unreactive in the plume. Surprisingly, Fe also appears to show little reactivity; the pronounced flocculation removal of Fe frequently observed in other estuaries is not seen in this system. This difference may be a consequence of the alkaline nature of the Mississippi which results in low dissolved Fe concentrations in the river (<50 nmol/kg). Zinc, another particle-reactive element, also shows little reactivity. This lack of reactivity for Zn, as well as Cu and Ni, is partly a result of the short residence time of plume waters in shallow areas affected by sedimentary interactions. The chromium distribution shows apparent non-conservative behavior indicative of estuarine removal; however, temporal variation in river concentrations is a more likely explanation for this behavior. For some other elements, complex distributions occur as a consequence of the interplay of physical-chemical and/or biological processes with the dynamic mixing regime. For Cd, desorption from the suspended load plays a major role in determining the distribution. However, sedimentary input may also play a role in the spatial variability of Cd. For V, biological uptake in the plume exerts a strong influence on its distribution. At the time of this study, uptake was large enough to consume both the river flux of V as well as a substantial amount of vanadium supplied by the ocean.

  15. Finite Element Modeling of the Behavior of Armor Materials Under High Strain Rates and Large Strains

    NASA Astrophysics Data System (ADS)

    Polyzois, Ioannis

    For years high strength steels and alloys have been widely used by the military for making armor plates. Advances in technology have led to the development of materials with improved resistance to penetration and deformation. Until recently, the behavior of these materials under high strain rates and large strains has been primarily based on laboratory testing using the Split Hopkinson Pressure Bar apparatus. With the advent of sophisticated computer programs, computer modeling and finite element simulations are being developed to predict the deformation behavior of these metals for a variety of conditions similar to those experienced during combat. In the present investigation, a modified direct impact Split Hopkinson Pressure Bar apparatus was modeled using the finite element software ABAQUS 6.8 for the purpose of simulating high strain rate compression of specimens of three armor materials: maraging steel 300, high hardness armor (HHA), and aluminum alloy 5083. These armor materials, provided by the Canadian Department of National Defence, were tested at the University of Manitoba by others. In this study, the empirical Johnson-Cook visco-plastic and damage models were used to simulate the deformation behavior obtained experimentally. A series of stress-time plots at various projectile impact momenta were produced and verified by comparison with experimental data. The impact momentum parameter was chosen rather than projectile velocity to normalize the initial conditions for each simulation. Phenomena such as the formation of adiabatic shear bands caused by deformation at high strains and strain rates were investigated through simulations. It was found that the Johnson-Cook model can accurately simulate the behavior of body-centered cubic (BCC) metals such as steels. The maximum shear stress was calculated for each simulation at various impact momenta. The finite element model showed that shear failure first occurred in the center of the cylindrical specimen and

  16. Recycle of Zirconium from Used Nuclear Fuel Cladding: A Major Element of Waste Reduction

    SciTech Connect

    Collins, Emory D; DelCul, Guillermo D; Terekhov, Dmitri; Emmanuel, N. V.

    2011-01-01

    Feasibility tests were initiated to determine if the zirconium in commercial used nuclear fuel (UNF) cladding can be recovered in sufficient purity to permit re-use, and if the recovery process can be operated economically. Initial tests are being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Early results indicate that quantitative recovery can be accomplished and product contamination with alloy constituents can be controlled sufficiently to meet purification requirements. Future tests with actual radioactive UNF cladding are planned. The objective of current research is to determine the feasibility of recovery and recycle of zirconium from used fuel cladding wastes. Zircaloy cladding, which contains 98+% of hafnium-free zirconium, is the second largest mass, on average {approx}25 wt %, of the components in used U.S. light-water-reactor fuel assemblies. Therefore, recovery and recycle of the zirconium would enable a large reduction in geologic waste disposal for advanced fuel cycles. Current practice is to compact or grout the cladding waste and store it for subsequent disposal in a geologic repository. This paper describes results of initial tests being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Future tests with actual radioactive UNF cladding are planned.

  17. Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix.

    SciTech Connect

    Kim, Y.S.; Hofman, G.

    2012-06-01

    Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

  18. Study on the mechanism of diametral cladding strain and mixed-oxide fuel element breaching in slow-ramp extended overpower transients

    SciTech Connect

    Tomoyuki Uwaba; Seiichiro Maeda; Tomoyasu Mizuno; Melissa C. Teague

    2012-10-01

    Cladding strain caused by fuel/cladding mechanical interaction (FCMI) was evaluated for mixed-oxide fuel elements subjected to 70–90% slow-ramp extended overpower transient tests in the experimental breeder reactor II. Calculated transient-induced cladding strains were correlated with cumulative damage fractions (CDFs) using cladding strength correlations. In a breached high-smeared density solid fuel element with low strength cladding, cladding thermal creep strain was significantly increased to approximately half the transient-induced cladding strain that was considered to be caused by the tertiary creep when the CDF was close to the breach criterion (=1.0), with the remaining strain due to instantaneous plastic deformation. In low-smeared density annular fuel elements, FCMI load was significantly mitigated and resulted in little cladding strain. The CDFs of the annular fuel elements were lower than 0.01 at the end of the overpower transient, indicating a substantial margin to breach. A substantial margin to breach was also maintained in a high-smeared density fuel element with high strength cladding.

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

    SciTech Connect

    Zhou, Wenzhong

    2011-03-24

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

  20. Analysis of the leaching behavior of elements from coal combustion residues for better management.

    PubMed

    Kumar, Ashvani; Samadder, S R

    2015-06-01

    In this study, fly ash, pond ash, bottom ash, slurry ash, raw water, pond water, and slurry samples were collected from Bokaro Thermal Power Plant, Bokaro, Jharkhand, India, and studied for the leachability of different elements by acid digestion and shake test at different liquid to solid (L/S) ratios. The raw water, pond water, slurry water, and leachates of acid digestion and shake tests were analyzed for the elements sodium (Na), potassium (K), calcium (Ca), iron (Fe), copper (Cu), cobalt (Co), manganese (Mn), cadmium (Cd), zinc (Zn), lead (Pb), nickel (Ni), and chromium (Cr). Shake test results confirmed that the water got saturated when the L/S ratio was equal to or above 10, indicating no further increase in concentration of elements at the L/S ratio of 10. Leaching behavior of Na from pond ash was not understood in the present study. In the study, the chemical composition showed that all the four types of ashes contain a small fraction of CaO (about 0.37 to 0.90 % by weight) and very high contents of SiO2 (about 55.14 to 58.34 % by weight) and Al2O3 (about 29.44 to 32.81 % by weight) that are the major composition of Portland cement. The study will help to understand the leachability potential of harmful elements present in fly ash, pond ash, bottom ash, and slurry ash under natural conditions and to take regulatory measures to protect the surface water, groundwater, and soil environment.

  1. Analysis of the leaching behavior of elements from coal combustion residues for better management.

    PubMed

    Kumar, Ashvani; Samadder, S R

    2015-06-01

    In this study, fly ash, pond ash, bottom ash, slurry ash, raw water, pond water, and slurry samples were collected from Bokaro Thermal Power Plant, Bokaro, Jharkhand, India, and studied for the leachability of different elements by acid digestion and shake test at different liquid to solid (L/S) ratios. The raw water, pond water, slurry water, and leachates of acid digestion and shake tests were analyzed for the elements sodium (Na), potassium (K), calcium (Ca), iron (Fe), copper (Cu), cobalt (Co), manganese (Mn), cadmium (Cd), zinc (Zn), lead (Pb), nickel (Ni), and chromium (Cr). Shake test results confirmed that the water got saturated when the L/S ratio was equal to or above 10, indicating no further increase in concentration of elements at the L/S ratio of 10. Leaching behavior of Na from pond ash was not understood in the present study. In the study, the chemical composition showed that all the four types of ashes contain a small fraction of CaO (about 0.37 to 0.90 % by weight) and very high contents of SiO2 (about 55.14 to 58.34 % by weight) and Al2O3 (about 29.44 to 32.81 % by weight) that are the major composition of Portland cement. The study will help to understand the leachability potential of harmful elements present in fly ash, pond ash, bottom ash, and slurry ash under natural conditions and to take regulatory measures to protect the surface water, groundwater, and soil environment. PMID:26002341

  2. Elemental Composition of Primary Aerosols Emitted from Burning of 21 Biomass Fuels Measured by Aerosol Mass Spectrometer

    NASA Astrophysics Data System (ADS)

    Desyaterik, Y.; Mack, L.; Lee, T.; Kreidenweis, S. M.; Collett, J. L.; Jimenez, J. L.; Worsnop, D. R.

    2010-12-01

    Biomass burning emissions are an important contributor to regional aerosol loading and have a large impact of on air quality, visibility, and radiative forcing. However, the detailed chemical composition of the aerosols emitted during biomass burning is largely unknown. In order to gain a better understanding of the chemical and physical properties of these emissions, 92 burns were undertaken in the combustion chamber of the USDA/FS Fire Sciences Laboratory in Missoula, Montana, in well-defined laboratory conditions. A set of 21 different fuels was tested that represents biomass burned annually in the western and southeastern U.S. The chemical composition of the resulting biomass smoke aerosols was analyzed with a high-resolution aerosol mass spectrometer (Aerodyne HR-ToF-AMS). Simultaneous measurements of CO2 and CO concentrations allowed flaming and smoldering fire regimes to be distinguished. The elemental composition of the organic portion of the aerosols was extracted from the AMS measurements. Here we present the variation of O/C, H/C and organic mass to organic carbon ratios (OM/OC) versus fire regime and fuel type. We also discuss the influence on the organic aerosol chemical composition of various factors such as fuel moisture content and total aerosol loading, as well as the approach used to account for water vapor ions derived from water originally present in sampled particles versus water vapor ions produced by electron impact fragmentation of organic molecules.

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

    SciTech Connect

    Hakan Ozaltun; Herman Shen; Pavel Madvedev

    2010-11-01

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

  4. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    NASA Astrophysics Data System (ADS)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

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

    SciTech Connect

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

    1987-10-01

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

  6. Predicting competitive adsorption behavior of major toxic anionic elements onto activated alumina: a speciation-based approach.

    PubMed

    Su, Tingzhi; Guan, Xiaohong; Tang, Yulin; Gu, Guowei; Wang, Jianmin

    2010-04-15

    Toxic anionic elements such as arsenic, selenium, and vanadium often co-exist in groundwater. These elements may impact each other when adsorption methods are used to remove them. In this study, we investigated the competitive adsorption behavior of As(V), Se(IV), and V(V) onto activated alumina under different pH and surface loading conditions. Results indicated that these anionic elements interfered with each other during adsorption. A speciation-based model was developed to quantify the competitive adsorption behavior of these elements. This model could predict the adsorption data well over the pH range of 1.5-12 for various surface loading conditions, using the same set of adsorption constants obtained from single-sorbate systems. This model has great implications in accurately predicting the field capacity of activated alumina under various local water quality conditions when multiple competitive anionic elements are present.

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

    DOEpatents

    Heckman, R.A.

    1980-12-19

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

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

    SciTech Connect

    J. L. Kloosterman; Abderrafi M. Ougouag

    2005-01-01

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

  9. Explorative First-Principles Simulation Study of Mineral - Melt Trace Element Partitioning Behavior

    NASA Astrophysics Data System (ADS)

    Wagner, J.; Jahn, S.

    2014-12-01

    Knowledge of trace element partition coefficients is crucial for a range of geoscientific applications. Obtaining the necessary experimental data is still a challenging and time consuming task, as the relevant processes typically take place under extreme P/T conditions. In this study, we explore a molecular scale simulation approach to predict mineral-melt partitioning. We use first principles molecular dynamics to investigate the rare earth element Y in the system garnet - melt, with focus on the influence of the melt. To predict the free energy change of the exchange reaction when Y is distributed between two phases, the method of thermodynamic integration is employed. Here we use an alchemical transmutation by which the identity (here expressed by its pseudopotential parameters) of a major element is gradually changed, in our case from 100% Al to 100% Y. The free energy change in turn enables us to predict the phase Y will partition into, as has been done previously for a CaO-Al2O3-SiO2-Y2O3 model system, employing classical force field simulations1. A major advantage of a molecular dynamics approach is that simulations contain information about the melt structure itself, thus enabling us to link observations to structural changes, e.g. a shift of average Y-O coordination number as a function of network connectivity. We choose a garnet-rich peridotite and a mid-ocean ridge basalt as starting compositions. After both melts are equilibrated at 3000 K and ambient pressure, we perform the thermodynamic integration and compare the free energy of the exchange reaction. The first results suggest that it is possible to at least qualitatively predict the behavior of Y in the two model systems, as compared to experimental findings. We will discuss the potential of the method to make quantitative predictions and how the effect of P and T can be evaluated. Potentially, this new tool may allow us to make predictions for almost any composition and condition available to

  10. Swelling behavior detection of irradiated U-10Zr alloy fuel using indirect neutron radiography

    NASA Astrophysics Data System (ADS)

    Sun, Yong; Huo, He-yong; Wu, Yang; Li, Jiangbo; Zhou, Wei; Guo, Hai-bing; Li, Hang; Cao, Chao; Yin, Wei; Wang, Sheng; Liu, Bin; Feng, Qi-jie; Tang, Bin

    2016-11-01

    It is hopeful that fusion-fission hybrid energy system will become an effective approach to achieve long-term sustainable development of fission energy. U-10Zr alloy (which means the mass ratio of Zr is 10%) fuel is the key material of subcritical blanket for fusion-fission hybrid energy system which the irradiation performance need to be considered. Indirect neutron radiography is used to detect the irradiated U-10Zr alloy because of the high residual dose in this paper. Different burnup samples (0.1%, 0.3%, 0.5% and 0.7%) have been tested with a special indirect neutron radiography device at CMRR (China Mianyang Research Reactor). The resolution of the device is better than 50 μm and the quantitative analysis of swelling behaviors was carried out. The results show that the swelling behaviors relate well to burnup character which can be detected accurately by indirect neutron radiography.

  11. Short-time scale behavior modeling within long-time scale fuel cycle evaluations

    SciTech Connect

    Johnson, M.; Tsvetkov, P.; Lucas, S.

    2012-07-01

    Typically, short-time and long-time scales in nuclear energy system behavior are accounted for with entirely separate models. However, long-term changes in system characteristics do affect short-term transients through material variations. This paper presents an approach to consistently account for short-time scales within a nuclear system lifespan. The reported findings and developments are of significant importance for small modular reactors and other nuclear energy systems operating in autonomous modes. It is necessary to simulate the short time-scale kinetic behavior of the reactor as well as the long time-scale dynamics that occur with fuel burnup. The former is modeled using the point kinetics equations, while the latter is modeled by the Bateman equations. (authors)

  12. Binary rare earth element-Ni/Co metallic glasses with distinct β-relaxation behaviors

    SciTech Connect

    Zhu, Z. G.; Wang, Z.; Wang, W. H.

    2015-10-21

    We report the formation of a series of rare earth element (RE)-Ni/Co binary metallic glasses (MGs) with unusual distinct β-relaxation peak compared with that of most of the reported MGs which usually exhibit as an excess wing or a shoulder. The β-relaxation behavior of RE-Ni/Co MGs is sensitive to the composition and the atomic radii of the RE and can be tuned through changing the fraction of RE-Ni (or Co) atomic pairs. The novel RE-Ni/Co MGs with distinct β-relaxation can serve as model system to investigate the nature of the β-relaxation as well as its relations with other physical and mechanical properties of MGs.

  13. Microstructure and corrosion behavior of binary titanium alloys with beta-stabilizing elements.

    PubMed

    Takada, Y; Nakajima, H; Okuno, O; Okabe, T

    2001-03-01

    Binary titanium alloys with the beta-stabilizing elements of Co, Cr, Cu, Fe, Mn and Pd (up to 30%) and Ag (up to 45%) were examined through metallographic observation and X-ray diffractometry to determine whether beta phases that are advantageous for dental use could be retained. Corrosion behavior was also investigated electrochemically and discussed thermodynamically. Some cast alloys with Co, Cr, Fe, Mn, and Pd retained the beta phase, whereas those with Ag and Cu had no beta phase. In some alloys, an intermetallic compound formed, based on information from the phase diagram. The corrosion resistance deteriorated in the TiAg alloys because Ti2Ag and/or TiAg intermetallic compounds preferentially dissolved in 0.9% NaCl solution. On the other hand, the remaining titanium alloys became easily passive and revealed good corrosion resistance similar to pure titanium since their matrices seemed to thermodynamically form titanium oxides as did pure titanium.

  14. Homogenized Finite Element Analysis on Effective Elastoplastic Mechanical Behaviors of Composite with Imperfect Interfaces

    PubMed Central

    Jiang, Wu-Gui; Zhong, Ren-Zhi; Qin, Qing H.; Tong, Yong-Gang

    2014-01-01

    A three-dimensional (3D) representative volume element (RVE) model was developed for analyzing effective mechanical behavior of fiber-reinforced ceramic matrix composites with imperfect interfaces. In the model, the fiber is assumed to be perfectly elastic until its tensile strength, and the ceramic material is modeled by an elasto-plastic Drucker-Prager constitutive law. The RVE model is then used to study the elastic properties and the tensile strength of composites with imperfect interfaces and validated through experiments. The imperfect interfaces between the fiber and the matrix are taken into account by introducing some cohesive contact surfaces. The influences of the interface on the elastic constants and the tensile strengths are examined through these interface models. PMID:25522170

  15. Mechanical behavior of aluminum-bearing ferritic alloys for accident-tolerant fuel cladding applications

    NASA Astrophysics Data System (ADS)

    Guria, Ankan

    Nuclear power currently provides about 13% of electrical power worldwide. Nuclear reactors generating this power traditionally use Zirconium (Zr) based alloys as the fuel cladding material. Exothermic reaction of Zr with steam under accident conditions may lead to production of hydrogen with the possibility of catastrophic consequences. Following the Fukushima-Daiichi incident, the exploration of accident-tolerant fuel cladding materials accelerated. Aluminum-rich (around 5 wt. %) ferritic steels such as Fecralloy, APMT(TM) and APM(TM) are considered as potential materials for accident-tolerant fuel cladding applications. These materials create an aluminum-based oxide scale protecting the alloy at elevated temperatures. Tensile deformation behavior of the above alloys was studied at different temperatures (25-500 °C) at a strain rate of 10-3 s-1 and correlated with microstructural characteristics. Higher strength and decent ductility of APMT(TM) led to further investigation of the alloy at various combination of strain rates and temperatures followed by fractography and detailed microscopic analyses. Serrations appeared in the stress-strain curves of APMT(TM) and Fecralloy steel tested in a limited temperature range (250-400 °C). The appearance of serrations is explained on the basis of dynamic strain aging (DSA) effect due to solute-dislocation interactions. The research in this study is being performed using the funds received from the US DOE Office of Nuclear Energy's Nuclear Energy University Programs (NEUP).

  16. Corrosion behavior of some high-temperature alloys under high velocity burnt fuel

    SciTech Connect

    Agarwal, D.C.; Brill, U.; Ibas, O.

    1995-12-31

    In a laboratory burner rig facility developed by Krupp VDM, the corrosion behavior of three high-temperature alloys was investigated under high velocity burnt fuel. A hot gas stream of burnt natural gas hits a sample at an angle of 45{degree}. Gas velocities of up to 80 m/s are obtained, and can be continuously adjusted by varying the air volume. By changing the sample to burner nozzle distance, a temperature gradient from 1,000 C in the center to 880 C at the edges of the sample can be achieved. Corrosion behavior of the two Fe-base alloys 310 S and 800H, and the Ni-base alloy 602CA, was evaluated by means of optical microscopy and SEM/EDAX analysis. According to results obtained so far, the alumina-former, alloy 602CA, provides best performance under high velocity burnt fuel at 880--1,000 C, as well as under steady state cyclic oxidation testing in air.

  17. Finite Element Modeling to Simulate the Elasto-Plastic Behavior of Polycrystalline in 718

    NASA Astrophysics Data System (ADS)

    Bonifaz, E. A.

    2013-01-01

    A 3D strain gradient plasticity finite element model was developed to simulate the elasto-plastic behavior of polycrystalline IN 718 alloys. The proposed model constructed in the basis of the so-called Kocks-Mecking model is used to determine the influence of microstructure attributes on the inelastic stress-strain distribution. Representative Volume Elements (RVEs) of different edge size but similar grain morphology and affordable computational meshes were tested to investigate the link between micro and macro variables of deformation and stress. The virtual specimens subjected to continuous monotonic straining loading conditions were constrained with random periodic boundary conditions. The difference in crystallographic orientation (which evolves in the process of straining) and the incompatibility of deformation between neighboring grains were accounted by the introduction of averaged Taylor factors and the evolution of geometrically necessary dislocation density. The effect of plastic deformation gradients imposed by the microstructure is clearly observed. Results demonstrate a strong dependence of flow stress and plastic strain on phase type and grain size. A main strategy for constitutive modeling of individual bulk grains is presented. The influence of the grain size on the aggregate response, in terms of local stress variations and aggregate elastic moduli was analyzed. It was observed that the elastic modulus in the bulk material is not dependent on grain size.

  18. Influence of multi-element ion beam bombardment on the corrosion behavior of iron and steel

    SciTech Connect

    Wei, Tian; Run, Wu; Weiping, Cai; Rutao, Wang ); Godechot, X.; Brown, I. )

    1991-06-01

    The effect of multi-element ion implantation on the corrosion resistance to acid solution has been studied for stainless steel, medium carbon steel, pure iron, and chromium-deposited iron. The implanted elements were Cu, Mo, Cr, Ni, Yb and Ti at doses of each species of from 5 {times} 10{sup 15} to 1 {times} 10{sup 17} cm{sup {minus}2} and at ion energies of up to 100 keV. The stainless steel used was 18-8 Cr-Ni, and the medium carbon steel was 0.45% C. The implanted samples were soaked in dilute sulfuric acid solution for periods up to 48 hours and the weight loss measured by atomic absorption spectroscopy. The kinetic parameter values describing the weight loss as a function of time were determined for all samples. In this paper we summarize the corrosion resistance behavior for the various different combinations of implanted species, doses, and substrates. The influence of the composition and structure of the modified surface layer is discussed.8 refs., 5 figs., 2 tabs.

  19. The effect of fuel thermal conductivity on the behavior of LWR cores during loss-of-coolant accidents

    SciTech Connect

    Terrani, Kurt A.; Wang, Dean; Ott, Larry J.; Montgomery, Robert O.

    2014-05-01

    The effect of variation in thermal conductivity of light water reactor fuel elements on core response during loss-of-coolant accident scenarios is examined. Initially, a simplified numerical analysis is utilized to determine the time scales associated with dissipation of stored energy from the fuel into the coolant once the fission reaction is stopped. The analysis is then followed by full reactor system thermal-hydraulics analysis of a typical boiling and pressurized water reactor subjected to a large break loss-of-coolant accident scenario using the TRACE code. Accordingly, sensitivity analyses to examine the effect of an increase in fuel thermal conductivity, up to 500%, on fuel temperature evolution during these transients are performed. Given the major differences in thermal-hydraulics design aspects of boiling and pressurized water reactors, different fuel and temperature responses during the simulated loss-of-coolant transients are observed.

  20. Partitioning behavior of aromatic components in jet fuel into diverse membrane-coated fibers.

    PubMed

    Baynes, Ronald E; Xia, Xin-Rui; Barlow, Beth M; Riviere, Jim E

    2007-11-01

    Jet fuel components are known to partition into skin and produce occupational irritant contact dermatitis (OICD) and potentially adverse systemic effects. The purpose of this study was to determine how jet fuel components partition (1) from solvent mixtures into diverse membrane-coated fibers (MCFs) and (2) from biological media into MCFs to predict tissue distribution. Three diverse MCFs, polydimethylsiloxane (PDMS, lipophilic), polyacrylate (PA, polarizable), and carbowax (CAR, polar), were selected to simulate the physicochemical properties of skin in vivo. Following an appropriate equilibrium time between the MCF and dosing solutions, the MCF was injected directly into a gas chromatograph/mass spectrometer (GC-MS) to quantify the amount that partitioned into the membrane. Three vehicles (water, 50% ethanol-water, and albumin-containing media solution) were studied for selected jet fuel components. The more hydrophobic the component, the greater was the partitioning into the membranes across all MCF types, especially from water. The presence of ethanol as a surrogate solvent resulted in significantly reduced partitioning into the MCFs with discernible differences across the three fibers based on their chemistries. The presence of a plasma substitute (media) also reduced partitioning into the MCF, with the CAR MCF system being better correlated to the predicted partitioning of aromatic components into skin. This study demonstrated that a single or multiple set of MCF fibers may be used as a surrogate for octanol/water systems and skin to assess partitioning behavior of nine aromatic components frequently formulated with jet fuels. These diverse inert fibers were able to assess solute partitioning from a blood substitute such as media into a membrane possessing physicochemical properties similar to human skin. This information may be incorporated into physiologically based pharmacokinetic (PBPK) models to provide a more accurate assessment of tissue dosimetry of

  1. Static and dynamic behavior of multiphase porous media: Governing equations and finite element implementation

    NASA Astrophysics Data System (ADS)

    Wei, Changfu

    2001-09-01

    The mechanical behavior of porous media such as geomaterials is largely governed by the interactions of the solid skeleton (or grains) with the fluids existing in the pores. These interactions occur through the interfaces between bulk components. Traditional analysis procedures of porous media, based on the principle of effective stress and Darcy's law, commonly fail to account for these interactions. In this dissertation, a continuum theory of multiphase porous media is developed, capable of rigorously characterizing the interactions among bulk components. Central to the theory is the implementation of the dynamic compatibility conditions that microscopically represent the constraints on the pressure jumps through interfaces. It is shown that Terzaghi's effective stress and capillary pressure can be characterized within a common framework. Within this context, a theoretical framework for poroelastoplasticity is developed, allowing the hysteresis in capillary pressure and plastic deformation of skeleton to be simulated in a hierarchical way. It is found that the mixture theory-based models of porous media can be linked with Biot's poroelasticity theory. A linear model based on the proposed theory is developed and used to analyze the propagation of acoustic waves in unsaturated soils and favorable comparisons to experimental results are obtained. A finite element procedure is developed and implemented into a computer code (called U_DYSAC2) for elastoplastic static and dynamic analyses of saturated and unsaturated porous media. Numerical examples including wave propagation, two-phase flow, consolidation, and seismic behavior of an embankment are presented. These examples show the capability of the theory for modeling a wide variety of behaviors of porous media.

  2. Electromechanical bending behavior study of soft photocurable ionogel actuator using a new finite element method

    NASA Astrophysics Data System (ADS)

    Wang, Zhipeng; He, Bin; Wang, Qigang; Yin, Yaobao

    2016-09-01

    The photocurable ionogel actuator (PIA) is one of the most promising driving mechanisms for the future due to its extraordinary features such as its light weight, flexibility, low-energy consumption and ability to work in open air. However, before the benefits of PIA can be effectively exploited for applications, a mathematical model is required to enhance the understanding of the parameters influencing the actuator electromechanical bending behavior. In this work, a model based on the finite element method (FEM) for the electromechanical bending behavior of PIA is established. It is assumed that the PIA consists of one ionogel layer and two activated carbon electrode layers. With reference to its operational principles, an analogy is drawn between thermal strain and induced strain in the PIA due to the volume change of the activated carbon electrode layer, which is a coupled structural/thermal model and can be solved by FEM. The distribution of net charge in the activated carbon electrode layer is mimicked using temperature distribution, and the electromechanical coupling coefficient is mimicked using the thermal expansion coefficient. Compared with the traditional equivalent bimorph beam model, the proposed model can predict the distribution of the induced strain more exactly. On the basis of the model, experiments are carried out to investigate the impact of selected parameters on the tip displacement, electromechanical coupling coefficient and induced strain of the PIA. The voltage of the input signal, and three geometrical parameters, length, width, and thickness, of the PIA are selected in this work. The experimental and simulation results indicate that the voltage, length, and thickness show significant influence on the electromechanical bending behavior of the PIA, but the width does not. As a whole, these results can be beneficial for providing enhanced degrees of understanding, predictability and control of PIA performance.

  3. Fuels and fire behavior dynamics on large-scale savanna fires in Kruger National Park, South Africa

    NASA Astrophysics Data System (ADS)

    Stocks, B. J.; van Wilgen, B. W.; Trollope, W. S. W.; McRae, D. J.; Mason, J. A.; Weirich, F.; Potgieter, A. L. F.

    1996-10-01

    Biomass characterization and fire behavior documentation were carried out on two large (>2000 ha) experimental fires conducted in arid savanna fuels in Kruger National Park in September 1992. Prefire fuel loads, fuel consumption, spread rates, flame zone characteristics, and in-fire and perimeter wind field dynamics were measured in order to determine overall energy release rates for each fire. Convection column dynamics were also measured in support of airborne trace gas and particulate measurements. Energy release rates varied significantly between the two fires, and this was strongly reflected in convection column development. The lower-intensity fire produced a weak, poorly defined smoke plume, while a well-developed column with a capping cumulus top developed during the higher intensity fire. Further experimental burning studies, in savannas with higher fuel loads, are recommended to further explore the fire behavior-convection column dynamics relationship investigated in this study.

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

    SciTech Connect

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

    2010-01-01

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

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

    SciTech Connect

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

    2007-12-01

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

  6. Behaviors of trace elements in Neoarchean and Paleoproterozoic paleosols: Implications for atmospheric oxygen evolution and continental oxidative weathering

    NASA Astrophysics Data System (ADS)

    Murakami, Takashi; Matsuura, Kei; Kanzaki, Yoshiki

    2016-11-01

    The behaviors of redox-sensitive and/or bio-essential trace elements in Neoarchean and Paleoproterozoic paleosols (ancient weathering profiles) were investigated to better understand atmospheric oxygen evolution. The loss or retention of individual trace elements in the paleosols can show how continental oxidative weathering, and thus atmospheric oxygen evolution, took place against age mainly due to their various redox potentials. The V, Cr, Ni, Cu, Zn and Mo concentrations of two Paleoproterozoic paleosols were measured by inductively coupled plasma optical emission spectrometry and mass spectrometry, and those, as well as Co, W and U concentrations, of nine Neoarchean and Paleoproterozoic paleosols were obtained from the literature. The trace element behaviors were constrained by their degrees of loss or retention in the paleosols. We applied two methods to the estimation: (i) retention fraction of element M (a mass ratio of element M of paleosol to parent rock using immobile elements such as Ti) and (ii) element-element (in particular, Si-element) correlations at different profile depths of a paleosol. There are two distinct groups in trace element behavior in the Neoarchean and Paleoproterozoic paleosols: Co, Ni, Zn and W were lost from weathering profiles until the early Paleoproterozoic but retained in the middle and late Paleoproterozoic, while V, Cr, Cu, Mo and U were retained in the profiles until the early Paleoproterozoic or slightly later but lost from the profiles in the middle and late Paleoproterozoic. More precisely, the timings of such loss and retention were different between trace elements during the Paleoproterozoic. The characteristics of these changes from retention to loss or from loss to retention indicate that the changes occurred and lasted throughout the Paleoproterozoic. The trace element behaviors, accordingly, suggest that continental weathering became oxidative progressively with age during almost the whole Paleoproterozoic, and thus

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

    NASA Technical Reports Server (NTRS)

    Walton, J. T.

    1994-01-01

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

  8. 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.

  9. Behavior of nuclear waste elements during hydrothermal alteration of glassy rhyolite in an active geothermal system: Yellowstone National Park, Wyoming

    SciTech Connect

    Sturchio, N.C.; Seitz, M.G.

    1984-12-31

    The behavior of a group of nuclear waste elements (U, Th, Sr, Zr, Sb, Cs, Ba, and Sm) during hydrothermal alteration of glassy rhyolite is investigated through detailed geochemical analyses of whole rocks, glass and mineral separates, and thermal waters. Significant mobility of U, Sr, Sb, Cs, and Ba is found, and the role of sorption processes in their observed behavior is identified. Th, Zr, and Sm are relatively immobile, except on a microscopic scale. 9 references, 2 figures, 2 tables.

  10. Americium and plutonium release behavior from irradiated mixed oxide fuel during heating

    NASA Astrophysics Data System (ADS)

    Sato, I.; Suto, M.; Miwa, S.; Hirosawa, T.; Koyama, S.

    2013-06-01

    The release behavior of Pu and Am was investigated under the reducing atmosphere expected in sodium cooled fast reactor severe accidents. Irradiated Pu and U mixed oxide fuels were heated at maximum temperatures of 2773 K and 3273 K. EPMA, γ-ray spectrometry and α-ray spectrometry for released and residual materials revealed that Pu and Am can be released more easily than U under the reducing atmosphere. The respective release rate coefficients for Pu and Am were obtained as 3.11 × 10-4 min-1 and 1.60 × 10-4 min-1 at 2773 K under the reducing atmosphere with oxygen partial pressure less than 0.02 Pa. Results of thermochemical calculations indicated that the main released chemical forms would likely be PuO for Pu and Am for Am under quite low oxygen partial pressure.

  11. Sliding Wear and Friction Behavior of Fuel Rod Material in Water and Dry State

    NASA Astrophysics Data System (ADS)

    Park, Jin Moo; Kim, Jae Hoon; Jeon, Kyeong Lak; Park, Jun Kyu

    In water cooled reactors, the friction between spacer grid and fuel rod can lead to severe wear and it is an important topic to study. In the present study, sliding wear behavior of zirconium alloy was investigated in water and dry state using the pin-on-disc sliding wear tester. Sliding wear resistance of zirconium alloy against heat treated inconel alloy was examined at room temperature. The parameters in this study were sliding velocity, axial load and sliding distance. The wear characteristics of zirconium alloy was evaluated by friction coefficient, specific wear rate and wear volume. The micro-mechanisms responsible for wear in zirconium alloy were identified to be micro-cutting, micro-pitting, delamination and micro-cracking of deformed surface zone.

  12. Investigation of silver and iodine transport through silicon carbide layers prepared for nuclear fuel element cladding

    NASA Astrophysics Data System (ADS)

    Friedland, E.; van der Berg, N. G.; Malherbe, J. B.; Hancke, J. J.; Barry, J.; Wendler, E.; Wesch, W.

    2011-03-01

    Transport of silver and iodine through polycrystalline SiC layers produced by PBMR (Pty) Ltd. for cladding of TRISO fuel kernels was investigated using Rutherford backscattering analysis and electron microscopy. Fluences of 2 × 10 16 Ag + cm -2 and 1 × 10 16 I + cm -2 were implanted at room temperature, 350 °C and 600 °C with an energy of 360 keV, producing an atomic density of approximately 1.5% at the projected ranges of about 100 nm. The broadening of the implantation profiles and the loss of diffusors through the front surface during vacuum annealing at temperatures up to 1400 °C was determined. The results for room temperature implantations point to completely different transport mechanisms for silver and iodine in highly disordered silicon carbide. However, similar results are obtained for high temperature implantations, although iodine transport is much stronger influenced by lattice defects than is the case for silver. For both diffusors transport in well annealed samples can be described by Fickian grain boundary diffusion with no abnormal loss through the surface as would be expected from the presence of nano-pores and/or micro-cracks. At 1100 °C diffusion coefficients for silver and iodine are below our detection limit of 10 -21 m 2 s -1, while they increase into the 10 -20 m 2 s -1 range at 1300 °C.

  13. Evaluating the manufacturability and combustion behaviors of sludge-derived fuel briquettes.

    PubMed

    Chiou, Ing-Jia; Wu, I-Tsung

    2014-10-01

    Based on the physical and chemical properties as well as calorific values of pulp sludge and textile sludge, this study investigates the differences between manufacturability, relationship between extrusion pressure and formability, as well as stability and combustion behaviors of extruded sludge-derived fuel briquettes (ESBB) and cemented sludge-derived fuel blocks (CSBB). The optimum proportion and relevant usage ESBB policies are proposed as well. Experimental results indicate that a large amount of water can be saved during the ESBB manufacturing process. Additionally, energy consumption decreases during the drying process. ESBB also has a more compact structure than that of CSBB, and its mean penetration loading is approximately 18.7 times higher as well. Moreover, the flame temperature of ESBB (624-968°C) is significantly higher than that of CSBB (393-517°C). Also, the dry bulk density and moisture regain of ESBB is significantly related to the penetration loading. Furthermore, the optimum mix proportion of ESBB is co-determined by the formability of pulp sludge and the calorific values of textile sludge. While considering the specific conditions (including formability, stability and calorific values), the recommended mix proportion for ESBB is PS50TS50. PMID:24913348

  14. Adsorption behavior of low concentration carbon monoxide on polymer electrolyte fuel cell anodes for automotive applications

    NASA Astrophysics Data System (ADS)

    Matsuda, Yoshiyuki; Shimizu, Takahiro; Mitsushima, Shigenori

    2016-06-01

    The adsorption behavior of CO on the anode around the concentration of 0.2 ppm allowed by ISO 14687-2 is investigated in polymer electrolyte fuel cells (PEFCs). CO and CO2 concentrations in the anode exhaust are measured during the operation of a JARI standard single cell at 60 °C cell temperature and 1000 mA cm-2 current density. CO coverage is estimated from the gas analysis and CO stripping voltammetry. The cell voltage decrease as a result of 0.2 ppm CO is 29 mV and the CO coverage is 0.6 at the steady state with 0.11 mg cm-2 of anode platinum loading. The CO coverage as a function of CO concentration approximately follows a Temkin-type isotherm. Oxygen permeated to the anode through a membrane is also measured during fuel cell operation. The exhaust velocity of oxygen from the anode was shown to be much higher than the CO supply velocity. Permeated oxygen should play an important role in CO oxidation under low CO concentration conditions.

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

    NASA Technical Reports Server (NTRS)

    Wang, Ten-See; Schutzenhofer, Luke

    1991-01-01

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

  16. Integrated remodeling-to-fracture finite element model of human proximal femur behavior.

    PubMed

    Hambli, Ridha; Lespessailles, Eric; Benhamou, Claude-Laurent

    2013-01-01

    The purpose of this work was to develop an integrated remodeling-to-fracture finite element model allowing for the combined simulation of (i) simulation of a human proximal femur remodeling under a given boundary conditions, (ii) followed by the simulation of its fracture behavior (force-displacement curve and fracture pattern) under quasi-static load. The combination of remodeling and fracture simulation into one unified model consists in considering that the femur properties resulting from the remodeling simulation correspond to the initial state for the fracture prediction. The remodeling model is based on phenomenological one based on a coupled strain and fatigue damage stimulus. The fracture model is based on continuum damage mechanics in order to predict the progressive fracturing process which allows to predict the fracture pattern and the complete force-displacement curve under quasi-static load. To prevent mesh-dependence that generally affects the damage propagation rate, regularization technique was applied in the current work. To investigate the potential of the proposed unified remodeling-to-fracture model, we performed remodeling simulations on a 3D proximal femur model for a duration of 365 days under five different daily loading conditions followed by a side fall fracture simulation reproducing previously published experimental tests (de Bakker et al. (2009), case C, male, 72 years old). We show here that the implementation of an integrated remodeling-to-fracture model provides more realistic prediction strategy to assess the bone remodeling effects on the fracture risk of bone.

  17. Low frequency finite element models of the acoustical behavior of earmuffs.

    PubMed

    Boyer, Sylvain; Doutres, Olivier; Sgard, Franck; Laville, Frédéric; Boutin, Jérôme

    2015-05-01

    This paper compares different approaches to model the vibroacoustic behavior of earmuffs at low frequency and investigates their accuracy by comparison with objective insertion loss measurements recently carried out by Boyer et al. [(2014). Appl. Acoust. 83, 76-85]. Two models based on the finite element (FE) method where the cushion is either modeled as a spring foundation (SF) or as an equivalent solid (ES), and the well-known lumped parameters model (LPM) are investigated. Modeling results show that: (i) all modeling strategies are in good agreement with measurements, providing that the characterization of the cushion equivalent mechanical properties are performed with great care and as close as possible to in situ loading, boundary, and environmental conditions and that the frequency dependence of the mechanical properties is taken into account, (ii) the LPM is the most simple modeling strategy, but the air volume enclosed by the earmuff must be correctly estimated, which is not as straightforward as it may seem, (iii) similar results are obtained with the SF and the ES FE-models of the cushion, but the SF should be preferred to predict the earmuff acoustic response at low frequency since it requires less parameters and a less complex characterization procedure. PMID:25994693

  18. Impact Behavior of A356 Foundry Alloys in the Presence of Trace Elements Ni and V

    NASA Astrophysics Data System (ADS)

    Casari, Daniele; Ludwig, Thomas H.; Merlin, Mattia; Arnberg, Lars; Garagnani, Gian Luca

    2015-02-01

    In the present work, the impact behavior of unmodified A356 alloys with the addition of Ni or V in as-cast and T6 heat-treated conditions was assessed. Charpy V-notched specimens obtained from sand and permanent mold casting showed low total absorbed energy average values ( W t < 2 J). SEM analysis of fracture profiles and surfaces indicated a Si-driven crack propagation with a predominant transgranular fracture mode. Occasionally, intergranular contributions to fracture were detected in the permanent mold cast alloys due to the locally finer microstructure. Concurrent mechanisms related to the chemical composition, solidification conditions and heat treatment were found to control the impact properties of the alloys. While the trace element Ni exerted only minor effects on the impact toughness of the A356 alloy, V had a strong influence: (i) V-containing sand cast alloys absorbed slightly higher impact energies compared to the corresponding A356 base alloys; (ii) in the permanent mold cast alloys, V in solid solution led to a considerable loss of ductility, which in turn decreased the total absorbed energy.

  19. Finite-element analysis of biting behavior and bone stress in the facial skeletons of bats.

    PubMed

    Dumont, Elizabeth R; Piccirillo, Justin; Grosse, Ian R

    2005-04-01

    The wide range of dietary niches filled by modern mammals is reflected in morphological diversity of the feeding apparatus. Despite volumes of data on the biomechanics of feeding, the extent to which the shape of mammal skulls reflects stresses generated by feeding is still unknown. In addition to the feeding apparatus, the skull accommodates the structural needs of the sensory systems and brain. We turned to bats as a model system for separating optimization for masticatory loads from optimization for other functions. Because the energetic cost of flight increases with body mass, it is reasonable to suggest that bats have experienced selective pressure over evolutionary time to minimize mass. Therefore, the skulls of bats are likely to be optimized to meet functional demands. We investigate the hypothesis that there is a biomechanical link between biting style and craniofacial morphology by combining biting behavior and bite force data gathered in the field with finite-element (FE) analysis. Our FE experiments compared patterns of stress in the craniofacial skeletons within and between two species of bats (Artibeus jamaicensis and Cynopterus brachyotis) under routine and atypical loading conditions. For both species, routine loading produced low stresses in most of the skull. However, the skull of Artibeus was most resistant to loads applied via its typical biting style, suggesting a mechanical link between routine loading and skull form. The same was not true of Cynopterus, where factors other than feeding appear to have had a more significant impact on craniofacial morphology.

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

    PubMed

    Saqib, Naeem; Bäckström, Mattias

    2014-12-01

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

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

    PubMed

    Saqib, Naeem; Bäckström, Mattias

    2014-12-01

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

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

    SciTech Connect

    Grogan, Brandon R; Mihalczo, John T

    2009-01-01

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

  3. Sintering behavior of lanthanide-containing glass-ceramic sealants for solid oxide fuel cells

    SciTech Connect

    Goel, Ashutosh; Reddy, Allu Amarnath; Pascual, Maria J.; Gremillard, Laurent; Malchere, Annie; Ferreira, Jose M.

    2012-05-01

    This article reports on the influence of different lanthanides (La, Nd, Gd and Yb) on sintering behavior of alkaline-earth aluminosilicate glass-ceramics sealants for their application in solid oxide fuel cells (SOFC). All the glasses have been prepared by melt-quench technique. The in situ follow up of sintering behavior of glass powders has been done by high temperature - environmental scanning electron microscope (HT-ESEM) and hot-stage microscope (HSM) while the crystalline phase evolution and assemblage has been analyzed by x-ray diffraction (XRD) and scanning electron microscopy (SEM). All the glass compositions exhibit a glass-in-glass phase separation followed by two stage sintering resulting in well sintered glass powder compacts after heat treatment at 850 C for 1 h. Diopside (CaMgSi{sub 2}O{sub 6}) based phases constituted the major crystalline part in glass-ceramics followed by some minor phases. The increase in lanthanide content in glasses suppressed their tendency towards devitrification, thus, resulting in glass-ceramics with high amount of residual glassy phase (50-96 wt.%) which is expected to facilitate their self-healing behavior during SOFC operation. The electrical conductivity of the investigated glass-ceramics varied between (1.19 and 7.33) x 10{sup -7} S cm{sup -1} (750-800 C), and depended on the ionic field strength of lanthanide cations. Further experimentation with respect to the long term thermal and chemical stability of residual glassy phase under SOFC operation conditions along with high temperature viscosity measurements will be required in order to elucidate the potential of these glass-ceramics as self-healing sealants.

  4. A fundamental study of the oxidation behavior of SI primary reference fuels with propionaldehyde and DTBP as an additive

    NASA Astrophysics Data System (ADS)

    Johnson, Rodney

    In an effort to combine the benefits of SI and CI engines, Homogeneous Charge Compression Ignition (HCCI) engines are being developed. HCCI combustion is achieved by controlling the temperature, pressure, and composition of the fuel and air mixture so that autoignition occurs in proper phasing with the piston motion. This control system is fundamentally more challenging than using a spark plug or fuel injector to determine ignition timing as in SI and CI engines, respectively. As a result, this is a technical barrier that must be overcome to make HCCI engines applicable to a wide range of vehicles and viable for high volume production. One way to tailor the autoignition timing is to use small amounts of ignition enhancing additives. In this study, the effect of the addition of DTBP and propionaldehyde on the autoignition behavior of SI primary reference fuels was investigated. The present work was conducted in a new research facility built around a single cylinder Cooperative Fuels Research (CFR) octane rating engine but modified to run in HCCI mode. It focused on the effect of select oxygenated hydrocarbons on hydrocarbon fuel oxidation, specifically, the primary reference fuels n-heptane and iso-octane. This work was conducted under HCCI operating conditions. Previously, the operating parameters for this engine were validated for stable combustion under a wide range of operating parameters such as engine speeds, equivalence ratios, compression ratios and inlet manifold temperature. The stable operating range under these conditions was recorded and used for the present study. The major focus of this study was to examine the effect of the addition of DTBP or propionaldehyde on the oxidation behavior of SI primary reference fuels. Under every test condition the addition of the additives DTBP and propionaldehyde caused a change in fuel oxidation. DTBP always promoted fuel oxidation while propionaldehyde promoted oxidation for lower octane number fuels and delayed

  5. The coupled kinetics of grain growth and fission product behavior in nuclear fuel under degraded-core accident conditions

    NASA Astrophysics Data System (ADS)

    Rest, J.

    1985-04-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, and cesium release from (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests (performed at Oak Ridge National Laboratory) and (2) trace-irradiated LWR fuel during severe-fuel-damage (SFD) tests (performed in the PBF reactor in Idaho). A theory of grain boundary sweeping of gas bubbles has been included within the FASTGRASS-VFP formalism. This theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges, and provides a means of determining whether gas bubbles are caught up and moved along by a moving grain boundary or whether the grain boundary is only temporarily retarded by the bubbles and then breaks away. In addition, as FASTGRASS-VFP provides for a mechanistic calculation of ultra- and intergranular fission product behavior, the coupled calculation between fission gas behavior and grain growth is kinetically comprehensive. Results of the analyses demonstrate that intragranular fission product behavior during both types of tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. The effect of fuel oxidation by steam on fission product and grain growth behavior is also considered. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in high-burnup fuel are highlighted.

  6. The Behavior of Chalcophile and Siderophile Elements during Magmatic Differentiation as Observed in Kilauea Iki Lava Lake, Hawaii

    NASA Astrophysics Data System (ADS)

    Greaney, A. T.; Rudnick, R. L.; Helz, R. L.; Gaschnig, R. M.; Ash, R. D.; Piccoli, P. M.

    2015-12-01

    In 1959, Kilauea Iki Lava Lake formed as a single pulse of picritic lava ponded in a preexisting crater. The lava cooled and differentiated over the following decades, providing an excellent natural laboratory to study basaltic differentiation. Major element, trace element, and data for numerous isotope systems of both eruption and drill core samples have been previously published. In this study, twelve chalcophile and siderophile elements (V, Ga, Ge, Mo, Ag, Cd, In, Sn, Sb, W, Tl, and Bi) were measured in sixteen whole rock samples using standard addition solution ICP-MS, which provides high precision data for elements that were previously undetectable. Samples ranging from 26.9 to 2.4 wt% MgO show that all of these elements display incompatible behavior throughout the lake as they increase exponentially with decreasing MgO wt%. Olivine and chromite are the only phases varying significantly in abundance in samples from 27 to 7 wt% MgO. Ferro-diabasic segregation veins and other internal differentiates (5.8 to 2.4 wt% MgO) consist of augite, plagioclase, Fe-Ti oxides, and an immiscible Cu-Fe sulfide phase. These veins are significantly more enriched in the listed elements than are the olivine basalts. Several elements (Mo, Ag, Cd, Sn, Sb, Tl) are well correlated with Cu (R2>0.84), which is assumed to be chalcophile in this system. Germanium appears to follow Ti while Ga, In, W, and Bi display incompatible behavior but don't directly correlate with other elements. Vanadium shows overall incompatible behavior but is depleted in extremely differentiated samples, suggesting it is sequestered in a late stage fractionating phase. Molybdenum, Sb, Tl, and Sn are also very well correlated (R2>0.95) with several incompatible lithophile elements (REE, Ba, Hf, Nb, Ta, Th). This suggests their overall behavior in Kilauea Iki Lava Lake isn't controlled by any fractionating phase, including sulfides, and they may behave in a more lithophile manner.

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

    SciTech Connect

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

    1983-02-01

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

  8. Site Selection and Pseudo-Clustering Behaviors of Alloying Elements in Aluminum-Lean γ-TiAl Intermetallics

    NASA Astrophysics Data System (ADS)

    Aykol, Muratahan; Mekhrabov, Amdulla O.; Vedat Akdeniz, M.

    2010-02-01

    Site selection and pseudo-clustering behaviors of the various M alloying elements in Al-lean Ti50Al50- X M X ( X = 1, 2, 3, 4, and 5 at. pct) intermetallics have been investigated by means of the ordering energy-dependent and long-range-order forced fast Monte Carlo simulation method. The ordering energies have been calculated via pseudopotential approximation in the electronic theory of alloys up to the third coordination sphere (CS) taking the anisotropic nature of tetragonal L10-type structure of γ-TiAl into account. It was shown that the site occupation characteristics of the M alloying element atoms in γ-TiAl intermetallics are governed by the relative magnitude of partial ordering energies between Ti-M and Al-M atomic pairs. However, the sign of partial ordering energies of these atomic pairs at the first CS becomes important in determining the clustering behavior and controls the dissolution modes of alloying element atoms in the γ-TiAl matrix. The pseudo-clustering behavior of alloying elements reveals three dissolution modes, namely, random dissolution (mode I), planar clustering in two dimensions (mode II), and three-dimensional (3-D) clustering (mode III) of the M occupant atoms.

  9. Multidimensional Fuel Performance Code: BISON

    SciTech Connect

    2014-09-03

    BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficiently solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phase field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.

  10. Multidimensional Fuel Performance Code: BISON

    2014-09-03

    BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficientlymore » solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phase field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.« less

  11. Electrostatic precipitator collection efficiency and trace element emissions from co-combustion of biomass and recovered fuel in fluidized-bed combustion.

    PubMed

    Lind, Terttaliisa; Hokkinen, Jouni; Jokiniemi, Jorma K; Saarikoski, Sanna; Hillamo, Risto

    2003-06-15

    Particle and trace element emissions from energy production have continuously been subject to tightening regulations. At the same time, not enough is known on the effect of different combustion processes and different fuels and fuel mixtures on the particle characteristics and particle removal device operation. In this investigation, electrostatic precipitator fractional collection efficiency and trace metal emissions were determined experimentally at a 66 MW biomass-fueled bubbling fluidized-bed combustion plant. The measurements were carried out at the inlet and outlet of the two-field electrostatic precipitator (ESP) at the flue gas temperature of 130-150 degrees C. Two fuel mixtures were investigated: biomass fuel containing 70% wood residue and 30% peat and biomass with recovered fuel containing 70% wood residue, 18% peat, and 12% recovered fuel. The particle mass concentration at the ESP inlet was 510-1400 mg/Nm3. Particle emission at the ESP outlet was 2.3-6.4 mg/Nm3. Total ESP collection efficiency was 99.2-99.8%. Collection efficiency had a minimum in particle size range of 0.1-2 microm. In this size range, collection efficiency was 96-97%. The emission of the trace metals As, Cd, Co, Cr, Cu, Mn, Ni, Pb, Sb, Tl, and V was well below the regulation values set by EU directive for waste incineration and co-incineration.

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

    SciTech Connect

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

    1982-08-01

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

  13. Ab initio studies of atomic properties and experimental behavior of element 119 and its lighter homologs.

    PubMed

    Borschevsky, A; Pershina, V; Eliav, E; Kaldor, U

    2013-03-28

    Static dipole polarizabilities of element 119 and its singly charged cation are calculated, along with those of its lighter homologs, Cs and Fr. Relativity is treated within the 4-component Dirac-Coulomb formalism and electron correlation is included by the single reference coupled cluster approach with single, double, and perturbative triple excitations (CCSD(T)). Very good agreement with available experimental values is obtained for Cs, lending credence to the predictions for Fr and element 119. The atomic properties in group-1 are largely determined by the valence ns orbital, which experiences relativistic stabilization and contraction in the heavier elements. As a result, element 119 is predicted to have a relatively low polarizability (169.7 a.u.), comparable to that of Na. The adsorption enthalpy of element 119 on Teflon, which is important for possible future experimental studies of this element, is estimated as 17.6 kJ/mol, the lowest among the atoms considered here. PMID:23556718

  14. Ab initio studies of atomic properties and experimental behavior of element 119 and its lighter homologs

    NASA Astrophysics Data System (ADS)

    Borschevsky, A.; Pershina, V.; Eliav, E.; Kaldor, U.

    2013-03-01

    Static dipole polarizabilities of element 119 and its singly charged cation are calculated, along with those of its lighter homologs, Cs and Fr. Relativity is treated within the 4-component Dirac-Coulomb formalism and electron correlation is included by the single reference coupled cluster approach with single, double, and perturbative triple excitations (CCSD(T)). Very good agreement with available experimental values is obtained for Cs, lending credence to the predictions for Fr and element 119. The atomic properties in group-1 are largely determined by the valence ns orbital, which experiences relativistic stabilization and contraction in the heavier elements. As a result, element 119 is predicted to have a relatively low polarizability (169.7 a.u.), comparable to that of Na. The adsorption enthalpy of element 119 on Teflon, which is important for possible future experimental studies of this element, is estimated as 17.6 kJ/mol, the lowest among the atoms considered here.

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

    PubMed Central

    Vouk, V B; Piver, W T

    1983-01-01

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

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

    SciTech Connect

    Johnson, A.; Nash, K.L.

    2013-07-01

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

  17. Neutron Flux Interpolation with Finite Element Method in the Nuclear Fuel Cell Calculation using Collision Probability Method

    SciTech Connect

    Shafii, M. Ali; Su'ud, Zaki; Waris, Abdul; Kurniasih, Neny; Ariani, Menik; Yulianti, Yanti

    2010-12-23

    Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the k{sub eff} and other parameters are investigated.

  18. Tribological behavior of near-frictionless carbon coatings in high- and low-sulfur diesel fuels.

    SciTech Connect

    Alzoubi, M. F.; Ajayi, O. O.; Eryilmaz, O. L.; Ozturk, O.; Erdemir, A.; Fenske, G.

    2000-01-19

    The sulfur content in diesel fuel has a significant effect on diesel engine emissions, which are currently subject to environmental regulations. It has been observed that engine particulate and gaseous emissions are directly proportional to fuel sulfur content. With the introduction of low-sulfur fuels, significant reductions in emissions are expected. The process of sulfur reduction in petroleum-based diesel fuels also reduces the lubricity of the fuel, resulting in premature failure of fuel injectors. Thus, another means of preventing injector failures is needed for engines operating with low-sulfur diesel fuels. In this study, the authors evaluated a near-frictionless carbon (NFC) coating (developed at Argonne National Laboratory) as a possible solution to the problems associated with fuel injector failures in low-lubricity fuels. Tribological tests were conducted with NFC-coated and uncoated H13 and 52100 steels lubricated with high- and low- sulfur diesel fuels in a high-frequency reciprocating test machine. The test results showed that the NFC coatings reduced wear rates by a factor of 10 over those of uncoated steel surfaces. In low-sulfur diesel fuel, the reduction in wear rate was even greater (i.e., by a factor of 12 compared to that of uncoated test pairs), indicating that the NFC coating holds promise as a potential solution to wear problems associated with the use of low-lubricity diesel fuels.

  19. Behavior of composite/metal aircraft structural elements and components under crash type loads: What are they telling us

    NASA Technical Reports Server (NTRS)

    Carden, Huey D.; Boitnott, Richard L.; Fasanella, Edwin L.

    1990-01-01

    Failure behavior results are presented from crash dynamics research using concepts of aircraft elements and substructure not necessarily designed or optimized for energy absorption or crash loading considerations. To achieve desired new designs which incorporate improved energy absorption capabilities often requires an understanding of how more conventional designs behave under crash loadings. Experimental and analytical data are presented which indicate some general trends in the failure behavior of a class of composite structures which include individual fuselage frames, skeleton subfloors with stringers and floor beams but without skin covering, and subfloors with skin added to the frame-stringer arrangement. Although the behavior is complex, a strong similarity in the static and dynamic failure behavior among these structures is illustrated through photographs of the experimental results and through analytical data of generic composite structural models. It is believed that the similarity in behavior is giving the designer and dynamists much information about what to expect in the crash behavior of these structures and can guide designs for improving the energy absorption and crash behavior of such structures.

  20. 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.

  1. Analysis of particle behavior in High-Velocity Oxy-Fuel thermal spraying process

    NASA Astrophysics Data System (ADS)

    Katanoda, Hiroshi; Matsuo, Kazuyasu

    2003-08-01

    This paper analyzes the behavior of coating particle as well as the gas flow both of inside and outside the High-Velocity Oxy-Fuel (HVOF) thermal spraying gun by using quasi-one-dimensional analysis and numerical simulation. The HVOF gun in the present analysis is an axisymmetric convergent-divergent nozzle with the design Mach number of 2.0 followed by a straight passage called barrel. In the present analysis it is assumed that the influence of the particles injected in the gas flow is neglected, and the interaction between the particles is also neglected. The gas flow in the gun is assumed to be quasi-one-dimensional adiabatic flow. The velocity, temperature and density of gas in the jet discharged from the barrel exit are predicted by solving Navier-Stokes equations numerically. The particle equation of motion is numerically integrated using three-step Runge-Kutta method. The drag coefficient of the particle is calculated by linear interpolation of the experimental data obtained in the past. Particle mean temperature is calculated by using Ranz and Marchalls’ correlation for spherical particles. From the present analysis, the distributions of velocity and temperature of the coating particles flying inside and outside the HVOF gun are predicted.

  2. Early fission-gas behavior in oxide fuel: escape vs trapping

    SciTech Connect

    Cordoliani, V.; Olander, D.

    2007-07-01

    A Monte Carlo code has been developed to investigate the behavior of fission-gas on the grain boundaries of UO{sub 2} fuel in the early stages of irradiation. Gas atoms arriving at the grain boundary from the grain interior undergo a random walk until they meet one of four fates: 1. nucleate a bubble by jumping next to another atom; 2. trapped by an existing bubble; 3. escape at the termination of the grain boundary at a free surface; 4. re-solved by collision with a fission fragment intersecting the grain boundary. At issue is the relative importance of escape (release) and trapping by intergranular bubbles. The switch of the dominant mechanism occurs at a critical burnup when the rate of trapping equals the arrival rate of gas atoms on the grain boundary. The critical burnup depends on fission rate and temperature, but is always less than 0.003 MWd/kgU. A very low rate of fission-gas release persists beyond the critical burnup; gas atoms from the grain interior that arrive at the grain boundary immediately adjacent to the escape surface have a small but finite probability of avoiding trapping and instead are released. The code calculates the intergranular bubble-size distribution with attention paid to the effect of re-solution. (authors)

  3. Thermal-behavior study of chlorine released from composite refuse derived fuel.

    PubMed

    Song, Zhi-Wei; Lv, Yi-Bo; Tong, Long-Yan

    2009-08-01

    In order to reduce secondary pollution during the incineration of composite refuse derived fuel (CRDF), the combustion features and the emission behavior of chlorine in CRDF containing coal were analyzed. The former was analyzed using thermo-gravimetric and the latter by gas chromatography-mass spectrometry. The release rate of inorganic chlorine during combustion reached 90 mass% at temperature between 773.15 and 873.15K. On the other hand, approximately 84 mass% release rates was resulting from pyrolysis at 723.15K. When temperature reached above 1073.15K, it was noticed that higher concentration of organic chlorine in different organic compounds were produced in the processing of pyrolysis compared with those released from the combustion processing. From the thermo-gravimetric analysis using a self-designed system, three distinct phases were detected in the thermal process of CRDF. The first phase occurred at temperature between 473 and 573K and its mass loss was about 38.50%. The second phase between temperature regions of 673-773K with a mass loss of 20.35%. The third phase was observed at the temperature between 873 and 1073K with 22.25% mass loss.

  4. Liquid Fuels: Pyrolytic Degradation and Fire Spread Behavior as Influenced by Buoyancy

    NASA Technical Reports Server (NTRS)

    Yeboah, Yaw D.; Malbrue, Courtney; Savage, Melane; Liao, Bo; Ross, Howard D. (Technical Monitor)

    2001-01-01

    This work is being conducted by the Combustion and Emission Control Lab in the Engineering Department at Clark Atlanta University under NASA Grant No. NCC3-707. The work aims at providing data to supplement the ongoing NASA research activities on fire spread across liquid pools by providing flow visualization and velocity measurements especially in the gas phase and gas-liquid interface. The fabrication, installation, and testing were completed during this reporting period. The system shakedown and detailed quantitative measurements with High Speed Video and Particle Image Velocimetry (PIV) systems using butanol as fuel were performed. New and interesting results, not previously reported in the literature, were obtained from the experiments using a modified NASA tray and butanol as fuel. Three distinct flame spread regimes, as previously reported, were observed. These were the pseudo-uniform regime below 20 C, the pulsating regime between 22 and 30 C and the uniform regime above about 31 C. In the pulsating regime the jump velocity appeared to be independent of the pool temperature. However, the retreat velocity between jumps appeared to depend on the initial pool temperature. The flame retreated before surging forwards with increasing brightness. Previous literature reported this phenomenon only under microgravity conditions. However, we observed such behavior in our normal gravity experiments. Mini-pulsations behind the flame front were also observed. Two or three of these pulsations were observed within a single flame front pulsating time period. The velocity vector maps of the gas and liquid phases ahead, during, and behind the flame front were characterized. At least one recirculation cell was observed right below the flame front.The size of the liquid phase vortex (recirculation cell) below the flame front appeared to decrease with increasing initial pool temperature. The experiments also showed how multiple vortices developed in the liquid phase. A large

  5. The behavior of fuel-lean premixed flames in a standard flammability limit tube under controlled gravity conditions

    NASA Technical Reports Server (NTRS)

    Wherley, B. L.; Strehlow, R. A.

    1986-01-01

    Fuel-lean flames in methane-air mixtures from 4.90 to 6.20 volume percent fuel and propane-air mixtures from 1.90 to 3.00 volume percent fuel were studied in the vicinity of the limit for a variety of gravity conditions. The limits were determined and the behavior of the flames studied for one g upward, one g downward, and zero g propagation. Photographic records of all flammability tube firings were obtained. The structure and behavior of these flames were detailed including the variations of the curvature of the flame front, the skirt length, and the occurrence of cellular instabilities with varying gravity conditions. The effect of ignition was also discussed. A survey of flame speeds as a function of mixture strength was made over a range of lean mixture compositions for each of the fuels studied. The results were presented graphically with those obtained by other researchers. The flame speed for constant fractional gravity loadings were plotted as a function of gravity loadings from 0.0 up to 2.0 g's against flame speeds extracted from the transient gravity flame histories for corresponding gravity loadings. The effects of varying gravity conditions on the extinguishment process for upward and downward propagating flames were investigated.

  6. Health status of workers of a thermal power station exposed for prolonged periods to arsenic and other elements from fuel.

    PubMed

    Buchancová, J; Klimentová, G; Knizková, M; Mesko, D; Gáliková, E; Kubík, J; Fabianová, E; Jakubis, M

    1998-02-01

    The Nováky Power Station (NPS) has been using since 1953 as fuel coal with a high content of As and with a low content of other metals. This involves a constant risk for the workers as well as pollution of the surroundings. The authors described 16 cases of chronic As intoxication in NPS workers after 22.3 +/- 8.4 years of exposure (especially stokers, maintenance workers, boiler cleaners). Among clinical symptoms prevailed sensory and motor polyneuropathy (13 cases), pseudoneurasthenic syndrome (10 cases), toxic encephalopathy (6 cases) and nasal septum perforation (2 cases). After 1989 the intoxications with As did not occur any more due to technical measures and health protection of the workers. The authors present a review of actual results of clinical, haematological and biochemical investigations and tests for metals (AAS methods) in biological materials of workers at risk in NPS (n = 70), exposed on average for 11.9 +/- 0.5 years, of average age 35.91 +/- 1.7 years (mean +/- SE) and compared the results to a matched control group of blood donors not exposed to metals (n = 29). In NPS workers significantly lower Hb values, higher serum creatinine, higher serum beta 2-microglobulin, higher As content in hair as well as higher serum Mn and Pb concentrations compared with the C-group were found. The exposed group had significantly lower serum Se concentrations (0.64 +/- 0.025 mumol/l (mean +/- SE) compared to Se levels of persons from an adjacent district. The authors stress the necessity of individual evaluation of the metal concentrations in relation to clinical findings. With prolonged exposure the situation can become more urgent not only because of chronic poisoning but also because of the cancerogenic effects of these elements on the human organism.

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

    PubMed

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

    2016-09-01

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

  8. Health status of workers of a thermal power station exposed for prolonged periods to arsenic and other elements from fuel.

    PubMed

    Buchancová, J; Klimentová, G; Knizková, M; Mesko, D; Gáliková, E; Kubík, J; Fabianová, E; Jakubis, M

    1998-02-01

    The Nováky Power Station (NPS) has been using since 1953 as fuel coal with a high content of As and with a low content of other metals. This involves a constant risk for the workers as well as pollution of the surroundings. The authors described 16 cases of chronic As intoxication in NPS workers after 22.3 +/- 8.4 years of exposure (especially stokers, maintenance workers, boiler cleaners). Among clinical symptoms prevailed sensory and motor polyneuropathy (13 cases), pseudoneurasthenic syndrome (10 cases), toxic encephalopathy (6 cases) and nasal septum perforation (2 cases). After 1989 the intoxications with As did not occur any more due to technical measures and health protection of the workers. The authors present a review of actual results of clinical, haematological and biochemical investigations and tests for metals (AAS methods) in biological materials of workers at risk in NPS (n = 70), exposed on average for 11.9 +/- 0.5 years, of average age 35.91 +/- 1.7 years (mean +/- SE) and compared the results to a matched control group of blood donors not exposed to metals (n = 29). In NPS workers significantly lower Hb values, higher serum creatinine, higher serum beta 2-microglobulin, higher As content in hair as well as higher serum Mn and Pb concentrations compared with the C-group were found. The exposed group had significantly lower serum Se concentrations (0.64 +/- 0.025 mumol/l (mean +/- SE) compared to Se levels of persons from an adjacent district. The authors stress the necessity of individual evaluation of the metal concentrations in relation to clinical findings. With prolonged exposure the situation can become more urgent not only because of chronic poisoning but also because of the cancerogenic effects of these elements on the human organism. PMID:9524739

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

    PubMed

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

    2016-09-01

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

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

    SciTech Connect

    Saqib, Naeem Bäckström, Mattias

    2014-12-15

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

  11. Effects of Mountain Pine Beetle on Fuels and Expected Fire Behavior in Lodgepole Pine Forests, Colorado, USA

    PubMed Central

    Schoennagel, Tania; Veblen, Thomas T.; Negron, José F.; Smith, Jeremy M.

    2012-01-01

    In Colorado and southern Wyoming, mountain pine beetle (MPB) has affected over 1.6 million ha of predominantly lodgepole pine forests, raising concerns about effects of MPB-caused mortality on subsequent wildfire risk and behavior. Using empirical data we modeled potential fire behavior across a gradient of wind speeds and moisture scenarios in Green stands compared three stages since MPB attack (Red [1–3 yrs], Grey [4–10 yrs], and Old-MPB [∼30 yrs]). MPB killed 50% of the trees and 70% of the basal area in Red and Grey stages. Across moisture scenarios, canopy fuel moisture was one-third lower in Red and Grey stages compared to the Green stage, making active crown fire possible at lower wind speeds and less extreme moisture conditions. More-open canopies and high loads of large surface fuels due to treefall in Grey and Old-MPB stages significantly increased surface fireline intensities, facilitating active crown fire at lower wind speeds (>30–55 km/hr) across all moisture scenarios. Not accounting for low foliar moistures in Red and Grey stages, and large surface fuels in Grey and Old-MPB stages, underestimates the occurrence of active crown fire. Under extreme burning conditions, minimum wind speeds for active crown fire were 25–35 km/hr lower for Red, Grey and Old-MPB stands compared to Green. However, if transition to crown fire occurs (outside the stand, or within the stand via ladder fuels or wind gusts >65 km/hr), active crown fire would be sustained at similar wind speeds, suggesting observed fire behavior may not be qualitatively different among MPB stages under extreme burning conditions. Overall, the risk (probability) of active crown fire appears elevated in MPB-affected stands, but the predominant fire hazard (crown fire) is similar across MPB stages and is characteristic of lodgepole pine forests where extremely dry, gusty weather conditions are key factors in determining fire behavior. PMID:22272268

  12. Effects of mountain pine beetle on fuels and expected fire behavior in lodgepole pine forests, Colorado, USA.

    PubMed

    Schoennagel, Tania; Veblen, Thomas T; Negron, José F; Smith, Jeremy M

    2012-01-01

    In Colorado and southern Wyoming, mountain pine beetle (MPB) has affected over 1.6 million ha of predominantly lodgepole pine forests, raising concerns about effects of MPB-caused mortality on subsequent wildfire risk and behavior. Using empirical data we modeled potential fire behavior across a gradient of wind speeds and moisture scenarios in Green stands compared three stages since MPB attack (Red [1-3 yrs], Grey [4-10 yrs], and Old-MPB [∼30 yrs]). MPB killed 50% of the trees and 70% of the basal area in Red and Grey stages. Across moisture scenarios, canopy fuel moisture was one-third lower in Red and Grey stages compared to the Green stage, making active crown fire possible at lower wind speeds and less extreme moisture conditions. More-open canopies and high loads of large surface fuels due to treefall in Grey and Old-MPB stages significantly increased surface fireline intensities, facilitating active crown fire at lower wind speeds (>30-55 km/hr) across all moisture scenarios. Not accounting for low foliar moistures in Red and Grey stages, and large surface fuels in Grey and Old-MPB stages, underestimates the occurrence of active crown fire. Under extreme burning conditions, minimum wind speeds for active crown fire were 25-35 km/hr lower for Red, Grey and Old-MPB stands compared to Green. However, if transition to crown fire occurs (outside the stand, or within the stand via ladder fuels or wind gusts >65 km/hr), active crown fire would be sustained at similar wind speeds, suggesting observed fire behavior may not be qualitatively different among MPB stages under extreme burning conditions. Overall, the risk (probability) of active crown fire appears elevated in MPB-affected stands, but the predominant fire hazard (crown fire) is similar across MPB stages and is characteristic of lodgepole pine forests where extremely dry, gusty weather conditions are key factors in determining fire behavior.

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

    NASA Technical Reports Server (NTRS)

    Broadway, Jeramie; Hickman, Robert; Mireles, Omar

    2012-01-01

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

  14. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    SciTech Connect

    Phillpot, Simon; Tulenko, James

    2011-09-08

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  15. Development and Behavior of Metallic Filter Element and Numerical Simulation of Transport Phenomena during Filter Regeneration Process

    SciTech Connect

    Kuang, C.; Zhang, J.; Wang, F.; Chen, J.

    2002-09-19

    Ceramic filters have revealed to have good thermal resistance and chemical corrosion resistance, but they are brittle and lack of toughness, and liable to rupture under large temperature swings. Metallic filters with their high strength and toughness and good heat conduction ability have showed good thermal shock resistance, 310S and FeAl intermetallic filter elements have exhibited additionally good chemical corrosion resistance in oxidizing and sulfidizing atmosphere( Sawada 1999 and Sunil et al. 1999). The behavior of metallic filter elements at high temperature was investigated and the filtration efficiency of the filter units for hot gas from a coal gasifier unit was tested. Pulse-jet cleaning of filter elements is a key component in the operation of the filtration unit. The pulse-jet is introduced into the filter element cavities from the clean side, and the dust cakes on the outer surfaces of the filter elements are detached and fall into the filter vessel. Sequential on-line cleaning of filter element groups yields a filter operation with no shutdown for filter regeneration. Development of advanced technologies in the design and operation of the pulse cleaning is one of the important tasks in order to increase the system reliability, to improve the filter life and to increase the filtering performance. The regeneration of filter element in gas filtration at high temperature plays a very important role for the operation of the process. Based on experimental observation and field operation, a numerical model is set up to numerically simulate the momentum and heat transport phenomena in the regeneration process, which is essential for understanding of the process, the optimization of process parameters and improvement of the design of the structure of venturi nozzle and the configuration of the apparatus.

  16. Experimental Investigation of Evaporation Behavior of Polonium and Rare-Earth Elements in Lead-Bismuth Eutectic Pool

    SciTech Connect

    Shuji Ohno; Shinya Miyahara; Yuji Kurata; Ryoei Katsura; Shigeru Yoshida

    2006-07-01

    Equilibrium evaporation behavior was experimentally investigated for polonium ({sup 210}Po) in liquid lead-bismuth eutectic (LBE) and for rare-earth elements gadolinium (Gd) and europium (Eu) in LBE to understand and clarify the transfer behavior of toxic impurities from LBE coolant to a gas phase. The experiments utilized the 'transpiration method' in which saturated vapor in an isothermal evaporation pot was transported by inert carrier gas and collected outside of the pot. While the previous paper ICONE12-49111 has already reported the evaporation behavior of LBE and of tellurium in LBE, this paper summarizes the outlines and the results of experiments for important impurity materials {sup 210}Po and rare-earth elements which are accumulated in liquid LBE as activation products and spallation products. In the experiments for rare-earth elements, non-radioactive isotope was used. The LBE pool is about 330-670 g in weight and has a surface area of 4 cm x 14 cm. {sup 210}Po experiments were carried out with a smaller test apparatus and radioactive {sup 210}Po produced through neutron irradiation of LBE in the Japan Materials Testing Reactor (JMTR). We obtained fundamental and instructive evaporation data such as vapor concentration, partial vapor pressure of {sup 210}Po in the gas phase, and gas-liquid equilibrium partition coefficients of the impurities in LBE under the temperature condition between 450 and 750 deg. C. The {sup 210}Po test revealed that Po had characteristics to be retained in LBE but was still more volatile than LBE solvent. A part of Eu tests implied high volatility of rare-earth elements comparable to that of Po. This tendency is possibly related to the local enrichment of the solute near the pool surface and needs to be investigated more. These results are useful and indispensable for the evaluation of radioactive materials transfer to the gas phase in LBE-cooled nuclear systems. (authors)

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

    NASA Technical Reports Server (NTRS)

    Klann, P. G.; Lantz, E.

    1973-01-01

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

  18. Investigation of the behavior of potentially hazardous trace elements in Kentucky coals and combustion byproducts

    SciTech Connect

    Robertson, J.D.; Blanchard, L.J.; Srikantapura, S.; Parekh, B.K.; Lafferty, C.J.

    1996-12-31

    The minor- and trace-element content of coal is of great interest because of the potentially hazardous impact on human health and the environment resulting from their release during coal combustion. Of the one billion tons of coal mined annually in the United States, 85-90% is consumed by coal-fired power plants. Potentially toxic elements present at concentrations as low as a few egg can be released in large quantities from combustion of this magnitude. Of special concern are those trace elements that occur naturally in coal which have been designated as potential hazardous air pollutants (HAPs) in the 1990 Amendments to the Clean Air Act. The principle objective of this work was to investigate a combination of physical and chemical coal cleaning techniques to remove 90 percent of HAP trace elements at 90 percent combustibles recovery from Kentucky No. 9 coal. Samples of this coal were first subjected to physical separation by flotation in a Denver cell. The float fraction from the Denver cell was then used as feed material for hydrothermal leaching tests in which the efficacy of dilute alkali (NaOH) and acid (HNO{sub 3}) solutions at various temperatures and pressures was investigated. The combined column flotation and mild chemical cleaning strategy removed 60-80% of trace elements with greater than 85, recovery of combustibles from very finely ground (-325 mesh) coal. The elemental composition of the samples generated at each stage was determined using particle induced X-ray emission (PIXE) analysis. PIXE is a rapid, instrumental technique that, in principle, is capable of analyzing all elements from sodium through uranium with sensitivities as low as 1 {mu}g/g.

  19. Behavior of fission gases in nuclear fuel: XAS characterization of Kr in UO2

    NASA Astrophysics Data System (ADS)

    Martin, P. M.; Vathonne, E.; Carlot, G.; Delorme, R.; Sabathier, C.; Freyss, M.; Garcia, P.; Bertolus, M.; Glatzel, P.; Proux, O.

    2015-11-01

    X-ray Absorption Spectroscopy (XAS) was used to study the behavior of krypton as a function of its concentration in UO2 samples implanted with Kr ions. For a 0.5 at.% krypton local concentration, by combining XAS results and DFT + U calculations, we show that without any thermal treatment Kr atoms are mainly incorporated in the UO2 lattice as single atoms inside a neutral bound Schottky defect with O vacancies aligned along the (100) direction (BSD1). A thermal treatment at 1273 K induces the precipitation of dense Kr nano-aggregates, most probably solid at room temperature. In addition, 26 ± 2% of the Kr atoms remain inside BSD1 showing that Kr-BSD1 complex is stable up to this temperature. Consequently, the (in-)solubility of krypton in UO2 has to be re-evaluated. For high Kr concentration (8 at.%), XAS signals show that Kr atoms have precipitated in nanometer-sized aggregates with internal densities ranging between 4.15(7) g cm-3 and 3.98(5) g cm-3 even after annealing at 873 K. By neglecting the effect due to the UO2 matrix, the corresponding krypton pressures at 300 K were equal to 2.6(3) GPa and 2.0(2) GPa, respectively. After annealing at 1673 K, regardless of the initial Kr concentration, a bi-modal distribution is observed with solid nano-aggregates even at room temperature and larger cavities only partially filled with Kr. These results are very close to those observed in UO2 fuel irradiated in reactor. In this study we show that a rare gas can be used as a probe to investigate the defect creation and their stability in UO2.

  20. Prediction of the start-up characteristics of a heat pipe-cooled thermionic fuel element (TFE)

    NASA Astrophysics Data System (ADS)

    Lieb, David; Witt, Tony; Lee, Celia; Miskolczy, Gabor; McVey, John

    A computer thermal model is used to predict the start-up characteristics of a heat pipe-cooled TFE. During start-up, the emitter temperature will increase, and heat will be radiated across the cesium gap to the collector/heat pipe and conducted from the top thermionic cell to the cesium-graphite reservoir through the emitter stem. A transient, finite element computer model of the top thermionic cell and the cesium-graphite reservoir was programmed to simulate the behavior of the collector heat pipe and reservoir. With the modeled cell configuration, the heat-choke coupling aids in heating the reservoir but is not extremely important. The calculation shows there is nearly enough direct heating of the collector-heat-pipe system to warm the TFE without requiring electron cooling. It is found that the thermal time constraints of the converter-reservoir system are well within 15 min and therefore will not be a limiting factor for rapid start-up of the reactor.

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

    PubMed

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

    2016-07-01

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

  2. Integrated study of the behavior of transuranic elements in the marine environment

    SciTech Connect

    Choppin, G. R.; Morse, J. W.

    1981-01-01

    In order to construct a model of radionuclide kinetics in an aquatic ecosystem, americium ions were chosen for study. Results will be applied to thorium, plutonium, neptunium and uranium for comparison of environmental behavior. (PSB)

  3. Elucidating the Effect of Alloying Elements on the Behavior of Austenitic Stainless Steels at Elevated Temperatures

    NASA Astrophysics Data System (ADS)

    Naghizadeh, Meysam; Mirzadeh, Hamed

    2016-09-01

    The effect of carbon and molybdenum on elevated temperature behavior of austenitic stainless steels was studied. It was revealed that carbon does not alter the overall grain coarsening behavior but molybdenum significantly retards the growth of grains toward higher temperatures and slower kinetics and effectively increases the grain growth activation energy due to an interaction energy between Mo and grain boundaries. These observations were based on especial activation energy plots, which facilitate the interpretation of results.

  4. A missing element in disaster mental health: behavioral health surveillance for first responders.

    PubMed

    Shubert, Jan; Ritchie, Elspeth Cameron; Everly, George S; Fiedler, Nancy; Williams, Mary Beth; Mitchell, Clifford S; Langlieb, Alan M

    2007-01-01

    Considerable literature exists on surveillance for medical effects of responses to a disaster but there is a dearth of information on conducting surveillance of behavioral health effects for first responders. This article reviews the literature and rationale behind behavioral health surveillance in the context of medical surveillance of first responders, examines special populations and ethical issues, discusses a model currently used by the U.S. military, discusses unresolved issues, and concludes with some practical suggestions.

  5. Finite element analysis of notch behavior using a state variable constitutive equation

    NASA Technical Reports Server (NTRS)

    Dame, L. T.; Stouffer, D. C.; Abuelfoutouh, N.

    1985-01-01

    The state variable constitutive equation of Bodner and Partom was used to calculate the load-strain response of Inconel 718 at 649 C in the root of a notch. The constitutive equation was used with the Bodner-Partom evolution equation and with a second evolution equation that was derived from a potential function of the stress and state variable. Data used in determining constants for the constitutive models was from one-dimensional smooth bar tests. The response was calculated for a plane stress condition at the root of the notch with a finite element code using constant strain triangular elements. Results from both evolution equations compared favorably with the observed experimental response. The accuracy and efficiency of the finite element calculations also compared favorably to existing methods.

  6. Adhesion and friction behavior of group 4 elements germanium, silicon, tin, and lead

    NASA Technical Reports Server (NTRS)

    Buckley, D. H.

    1975-01-01

    Adhesion and friction studies were conducted with thin films of the group IV elements silicon, germanium, tin, and lead ion plated on the nickel (011) substrate. The mating surface was gold (111). Contacts were made for the elements in the clean state and with oxygen present. Adhesion and friction experiments were conducted at very light loads of 1 to 10 g. Sliding was at a speed of 0.7 mm/min. Friction results indicate that the more covalently bonded elements silicon and germanium exhibit lower adhesion and friction than the more metallic bonded tin and lead. The adhesion of gold to germanium was observed, and recrystallization of the transferred gold occurred. Plastic flow of germanium was seen with sliding. Oxygen reduced, but did not eliminate, the adhesion observed with germanium and silicon.

  7. Representative volume element to estimate buckling behavior of graphene/polymer nanocomposite

    PubMed Central

    2012-01-01

    The aim of the research article is to develop a representative volume element using finite elements to study the buckling stability of graphene/polymer nanocomposites. Research work exploring the full potential of graphene as filler for nanocomposites is limited in part due to the complex processes associated with the mixing of graphene in polymer. To overcome some of these issues, a multiscale modeling technique has been proposed in this numerical work. Graphene was herein modeled in the atomistic scale, whereas the polymer deformation was analyzed as a continuum. Separate representative volume element models were developed for investigating buckling in neat polymer and graphene/polymer nanocomposites. Significant improvements in buckling strength were observed under applied compressive loading when compared with the buckling stability of neat polymer. PMID:22994951

  8. Atmospheric behavior of trace elements on particles emitted from a coal-fired power plant

    NASA Astrophysics Data System (ADS)

    Ondov, J. M.; Choquette, C. E.; Zoller, W. H.; Gordon, G. E.; Biermann, A. H.; Heft, R. E.

    Filter and cascade impactor samples of suspended particles were collected in-stack and at distances up to 64 km downwind in the plume of a large western coal-fired power plant equipped with both electrostatic precipitators (ESP) and venturi wet particulate scrubbers (VWS) to investigate modifications of the particulate signatures of minor and trace elements during transport. Samples were analyzed for 40 elements by instrumental neutron activation analysis. Precipitator malfunction during the experiment caused greater than normal emission of large particles, and concentrations of As, Zn, Sb, Mo, Ga, W, U, V and Ba in near-plume particles collected on filters were enriched relative to their concentrations in stack particles by factors of 1.4 to 2.5, presumably because of sedimentation of very large particles. Selenium was enriched by up to 6-fold (plume:stack). However, enrichment of elements in the plume relative to more typical in-stack particles were insignificant for all elements except Se, which was enriched 2.3-fold. Concentrations of Se on particles in the stack and plume suggest that most of the Se vapor in stack gases became associated with aerosol particles soon after emission. Thus although significant post-emission modifications of elemental signatures of particles may occur for poorly controlled plants, little change is expected for well-controlled plants equipped with ESPs except for Se. Source signatures measured for Se must account for vapor deposition. Impactor data showed a preferential decrease in the concentrations of the above elements in submicrometer particles; suggesting that either intermodal coagulation or size selective sampling losses were important. The impactor data further suggest that enrichment-particle-size profiles for VWS emissions were not conservative during transport.

  9. Corrosion Minimization for Research Reactor Fuel

    SciTech Connect

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  10. A Micromechanics Finite Element Model for Studying the Mechanical Behavior of Spray-On Foam Insulation (SOFI)

    NASA Technical Reports Server (NTRS)

    Ghosn, Louis J.; Sullivan, Roy M.; Lerch, Bradley A.

    2006-01-01

    A micromechanics model has been constructed to study the mechanical behavior of spray-on foam insulation (SOFI) for the external tank. The model was constructed using finite elements representing the fundamental repeating unit of the SOFI microstructure. The details of the micromechanics model were based on cell observations and measured average cell dimensions discerned from photomicrographs. The unit cell model is an elongated Kelvin model (fourteen-sided polyhedron with 8 hexagonal and six quadrilateral faces), which will pack to a 100% density. The cell faces and cell edges are modeled using three-dimensional 20-node brick elements. Only one-eighth of the cell is modeled due to symmetry. By exercising the model and correlating the results with the macro-mechanical foam behavior obtained through material characterization testing, the intrinsic stiffness and Poisson s Ratio of the polymeric cell walls and edges are determined as a function of temperature. The model is then exercised to study the unique and complex temperature-dependent mechanical behavior as well as the fracture initiation and propagation at the microscopic unit cell level.

  11. Mobility behavior and environmental implications of trace elements associated with coal gangue: a case study at the Huainan Coalfield in China.

    PubMed

    Chuncai, Zhou; Guijian, Liu; Dun, Wu; Ting, Fang; Ruwei, Wang; Xiang, Fan

    2014-01-01

    The potential environmental hazards posed by trace elements have assumed serious proportions due to their toxicity, bioavailability and geochemical behavior. The toxicity and mobility of trace elements in coal gangue is dependent on the elements' chemical properties, therefore, the quantification of the different forms of trace elements is more significant than the estimation of their total concentrations. In this study, the mobility behavior of trace elements in coal gangue from the Huainan Coalfield was studied to evaluate the potential eco-toxicity of the trace elements. Sequential extraction was employed to analyze the fractionation behavior of trace elements in coal gangue. The selected trace elements (As, Co, Cr, Cu, Mn, Ni, Se, Sn, V and Zn) are predominantly found in silicate-bound, sulfide-bound and carbonate-bound fractions. The correlation of the element concentration with ash yield, aluminum, calcium and iron-sulfur indicates that As, Co, Cu, Ni, Se and Zn in coal gangue are mainly associated with sulfide minerals, which could release from coal gangue easily and can disperse into the environment as a result of long-term natural weathering. The Risk Assessment Code reveals that the trace elements (Mn, Cr, Se, Ni, Zn, As and Cu) can pose serious environmental risks to the ecosystem. The fractionation profiles of other elements (Co, Sn and V) indicate no risk or low risk to the environment.

  12. Investigation into the diffusion and oxidation behavior of the interface between a plasma-sprayed anode and a porous steel support for solid oxide fuel cells

    NASA Astrophysics Data System (ADS)

    Zhang, Shan-Lin; Li, Cheng-Xin; Li, Chang-Jiu; Liu, Meilin; Yang, Guan-Jun

    2016-08-01

    Porous metal-supported solid oxide fuel cells (SOFCs) have attracted much attention because their potential to dramatically reduce the cost while enhancing the robustness and manufacturability. In particular, 430 ferritic steel (430L) is one of the popular choice for SOFC support because of its superior performance and low cost. In this study, we investigate the oxidation and diffusion behavior of the interface between a Ni-based anode and porous 430L support exposed to a humidified (3% H2O) hydrogen atmosphere at 700 °C. The Ni-GDC (Ce0.8Gd0.2O2-δ) cermet anodes are deposited on the porous 430L support by atmospheric plasma spraying (APS). The effect of exposure time on the microstructure and phase structure of the anode and the supports is studied and the element diffusion across the support/anode interface is characterized. Results indicate that the main oxidation product of the 430L support is Cr2O3, and that Cr and Fe will diffuse to the anode and the diffusion thickness increases with the exposure time. The diffusion thickness of Cr and Fe reach about 5 and 2 μm, respectively, after 1000 h exposure. However, the element diffusion and oxidation has little influence on the area-specific resistance, indicating that the porous 430L steel and plasma sprayed Ni-GDC anode are promising for durable SOFCs.

  13. Trace element emissions

    SciTech Connect

    Benson, S.A.; Erickson, T.A.; Steadman, E.N.; Zygarlicke, C.J.; Hauserman, W.B.; Hassett, D.J.

    1994-10-01

    The Energy & Environmental Research Center (EERC) is carrying out an investigation that will provide methods to predict the fate of selected trace elements in integrated gasification combined cycle (IGCC) and integrated gasification fuel cell (IGFC) systems to aid in the development of methods to control the emission of trace elements determined to be air toxics. The goal of this project is to identify the effects of critical chemical and physical transformations associated with trace element behavior in IGCC and IGFC systems. The trace elements included in this project are arsenic, chromium, cadmium, mercury, nickel, selenium, and lead. The research seeks to identify and fill, experimentally and/or theoretically, data gaps that currently exist on the fate and composition of trace elements. The specific objectives are to (1) review the existing literature to identify the type and quantity of trace elements from coal gasification systems, (2) perform laboratory-scale experimentation and computer modeling to enable prediction of trace element emissions, and (3) identify methods to control trace element emissions.

  14. Finite Element Analysis of 2.5D Woven Composites, Part II: Damage Behavior Simulation and Strength Prediction

    NASA Astrophysics Data System (ADS)

    Song, Jian; Wen, Weidong; Cui, Haitao; Zhang, Hongjian; Xu, Ying

    2016-02-01

    In the first part of the work, a new 2.5D woven composites finite element model (2.5D WCFEM) which took into consideration the impact of face structures and can accurately predict the main elastic performances has been established. In this part, the stress-strain behavior and the damage characteristic of this material under uniaxial tension are simulated using nonlinear progressive damage analysis based on damage mechanics. Meanwhile, experimental investigation and fracture analysis are conducted to evaluate the validity of the proposed method. Finally, the influence of woven parameters on the mechanical behavior is discussed. Compared with the test results, a good agreement between the computational and experimental results has been obtained. The progressive damage characteristic and main failure modes are also revealed.

  15. Abundances of Ag and Cu in mantle peridotites and the implications for the behavior of chalcophile elements in the mantle

    NASA Astrophysics Data System (ADS)

    Wang, Zaicong; Becker, Harry

    2015-07-01

    Silver abundances in mantle peridotites and the behavior of Ag during high temperature mantle processes have received little attention and, as a consequence, the abundance of Ag in the bulk silicate Earth (BSE) has been poorly constrained. In order to better understand the processes that fractionate Ag and other chalcophile elements in the mantle, abundances of Ag and Cu in mantle peridotites from different geological settings (n = 68) have been obtained by isotope dilution ICP-MS methods. In peridotite tectonites and in a few suites of peridotite xenoliths which display evidence for variable extents of melt depletion and refertilization by silicate melts, Ag and Cu abundances show positive correlations with moderately incompatible elements such as S, Se, Te and Au. The mean Cu/Ag in fertile peridotites (3500 ± 1200, 1s, n = 38) is indistinguishable from the mean Cu/Ag of mid ocean ridge basalts (MORB, 3600 ± 400, 1s, n = 338) and MORB sulfide droplets. The constant mean Cu/Ag ratios indicate similar behavior of Ag and Cu during partial melting of the mantle, refertilization and magmatic fractionation, and thus should be representative of the Earth's upper mantle. The systematic fractionation of Cu, Ag, Au, S, Se and Te in peridotites and basalts is consistent with sulfide melt-silicate melt partitioning with apparent partition coefficients of platinum group elements (PGE) > Au ⩾ Te > Cu ≈ Ag > Se ⩾ S. Because of the effects of secondary processes, the abundances of chalcophile elements, notably S, Se, but also Cu and the PGE in many peridotite xenoliths are variable and lower than in peridotite massifs. Refertilization of peridotite may change abundances of chalcophile and lithophile elements in peridotite massifs, however, this seems to mostly occur in a systematic way. Correlations with lithophile and chalcophile elements and the overlapping mean Cu/Ag ratios of peridotites and ocean ridge basalts are used to constrain abundances of Ag and Cu in the BSE

  16. Organic double layer element driven by triboelectric nanogenerator: Study of carrier behavior by non-contact optical method

    NASA Astrophysics Data System (ADS)

    Chen, Xiangyu; Taguchi, Dai; Manaka, Takaaki; Iwamoto, Mitsumasa

    2016-02-01

    By using optical electric-field-induced second-harmonic generation (EFISHG) technique, we studied carrier behavior caused by contact electrification (CE) in an organic double-layer element. This double-layer sample was half suspended in the open air, where one electrode (anode or cathode) was connected with a Cu foil for electrification while the other electrode was floated. Results showed two distinct carrier behaviors, depending on the (anode or cathode) connections to the Cu foil, and these carrier behaviors were analyzed based on the Maxwell-Wagner model. The double-layer sample works as a simple solar cell device. The photovoltaic effect and CE process have been proved to be two paralleled effects without strong interaction with each other, while photoconductivity changing in the sample can enhance the relaxation of CE induced charges. By probing the carrier behavior in this half-suspended device, the EFISHG technique has been demonstrated to be an effective non-contact method for clarifying the CE effect on related energy harvesting devices and electronics devices. Meanwhile, the related physical analysis in this letter is also useful for elucidating the fundamental characteristic of hybrid energy system based on solar cell and triboelectric nanogenerator.

  17. Microbiota Modulates Behavior and Protein Kinase C mediated cAMP response element-binding protein Signaling.

    PubMed

    Zeng, Li; Zeng, Benhua; Wang, Haiyang; Li, Bo; Huo, Ran; Zheng, Peng; Zhang, Xiaotong; Du, Xiangyu; Liu, Meiling; Fang, Zheng; Xu, Xuejiao; Zhou, Chanjuan; Chen, Jianjun; Li, Wenxia; Guo, Jing; Wei, Hong; Xie, Peng

    2016-01-01

    Evolutionary pressure drives gut microbiota-host coevolution and results in complex interactions between gut microbiota and neural development; however, the molecular mechanisms by which the microbiota governs host behavior remain obscure. Here, we report that colonization early in life is crucial for the microbiota to modulate brain development and behavior; later colonization or deletion of microbiota cannot completely reverse the behaviors. Microarray analysis revealed an association between absence of gut microbiota and expression in cAMP responding element-binding protein (CREB) regulated genes in the hippocampus. The absence of gut microbiota from birth was shown to be associated with decreased CREB expression, followed by decreases of protein kinase C beta (PRKCB) and AMPA receptors expression, and an increase of phosphorylation CREB (pCREB) expression. Microbiota colonization in adolescence restored CREB and pCREB expression, but did not alter PRKCB and AMPARs expression. The removal of the gut microbiota from SPF mice using antibiotics only reduced pCREB expression. These findings suggest that (i) colonization of the gut microbiota early in life might facilitate neurodevelopment via PKC-CREB signaling and (ii) although GF mice and ABX mice display reduced anxiety-related behaviors, the molecular mechanisms behind this might differ. PMID:27444685

  18. Microbiota Modulates Behavior and Protein Kinase C mediated cAMP response element-binding protein Signaling

    PubMed Central

    Zeng, Li; Zeng, Benhua; Wang, Haiyang; Li, Bo; Huo, Ran; Zheng, Peng; Zhang, Xiaotong; Du, Xiangyu; Liu, Meiling; Fang, Zheng; Xu, Xuejiao; Zhou, Chanjuan; Chen, Jianjun; Li, Wenxia; Guo, Jing; Wei, Hong; Xie, Peng

    2016-01-01

    Evolutionary pressure drives gut microbiota–host coevolution and results in complex interactions between gut microbiota and neural development; however, the molecular mechanisms by which the microbiota governs host behavior remain obscure. Here, we report that colonization early in life is crucial for the microbiota to modulate brain development and behavior; later colonization or deletion of microbiota cannot completely reverse the behaviors. Microarray analysis revealed an association between absence of gut microbiota and expression in cAMP responding element-binding protein (CREB) regulated genes in the hippocampus. The absence of gut microbiota from birth was shown to be associated with decreased CREB expression, followed by decreases of protein kinase C beta (PRKCB) and AMPA receptors expression, and an increase of phosphorylation CREB (pCREB) expression. Microbiota colonization in adolescence restored CREB and pCREB expression, but did not alter PRKCB and AMPARs expression. The removal of the gut microbiota from SPF mice using antibiotics only reduced pCREB expression. These findings suggest that (i) colonization of the gut microbiota early in life might facilitate neurodevelopment via PKC–CREB signaling and (ii) although GF mice and ABX mice display reduced anxiety-related behaviors, the molecular mechanisms behind this might differ. PMID:27444685

  19. Some recent observations on the radiation behavior of uranium silicide dispersion fuel

    SciTech Connect

    Hofman, G.L.

    1988-01-01

    Addition of B{sub 4}C burnable poison results in higher plate swelling in both U{sub 3}Si{sub 2} and U{sub 3}Si-Al dispersion fuel plates and also decreases the blister threshold temperature of these plates. Prolonged annealing of U{sub 3}Si{sub 2}-Al fuel plates produced no blister after 696 hours at 400{degrees}C. Blister formation started between 257 hours and 327 hours at 425{degrees}C and between 115 hours and 210 hours at 450{degrees}C. Operation with breached cladding resulted in pillowing of an U{sub 3}Si-Al fuel plate due to reaction of the fuel core with coolant water. 4 refs., 10 figs., 2 tabs.

  20. HIV-1 and M-PMV RNA Nuclear Export Elements Program Viral Genomes for Distinct Cytoplasmic Trafficking Behaviors

    PubMed Central

    Pocock, Ginger M.; Becker, Jordan T.; Swanson, Chad M.; Ahlquist, Paul; Sherer, Nathan M.

    2016-01-01

    Retroviruses encode cis-acting RNA nuclear export elements that override nuclear retention of intron-containing viral mRNAs including the full-length, unspliced genomic RNAs (gRNAs) packaged into assembling virions. The HIV-1 Rev-response element (RRE) recruits the cellular nuclear export receptor CRM1 (also known as exportin-1/XPO1) using the viral protein Rev, while simple retroviruses encode constitutive transport elements (CTEs) that directly recruit components of the NXF1(Tap)/NXT1(p15) mRNA nuclear export machinery. How gRNA nuclear export is linked to trafficking machineries in the cytoplasm upstream of virus particle assembly is unknown. Here we used long-term (>24 h), multicolor live cell imaging to directly visualize HIV-1 gRNA nuclear export, translation, cytoplasmic trafficking, and virus particle production in single cells. We show that the HIV-1 RRE regulates unique, en masse, Rev- and CRM1-dependent “burst-like” transitions of mRNAs from the nucleus to flood the cytoplasm in a non-localized fashion. By contrast, the CTE derived from Mason-Pfizer monkey virus (M-PMV) links gRNAs to microtubules in the cytoplasm, driving them to cluster markedly to the centrosome that forms the pericentriolar core of the microtubule-organizing center (MTOC). Adding each export element to selected heterologous mRNAs was sufficient to confer each distinct export behavior, as was directing Rev/CRM1 or NXF1/NXT1 transport modules to mRNAs using a site-specific RNA tethering strategy. Moreover, multiple CTEs per transcript enhanced MTOC targeting, suggesting that a cooperative mechanism links NXF1/NXT1 to microtubules. Combined, these results reveal striking, unexpected features of retroviral gRNA nucleocytoplasmic transport and demonstrate roles for mRNA export elements that extend beyond nuclear pores to impact gRNA distribution in the cytoplasm. PMID:27070420

  1. HIV-1 and M-PMV RNA Nuclear Export Elements Program Viral Genomes for Distinct Cytoplasmic Trafficking Behaviors.

    PubMed

    Pocock, Ginger M; Becker, Jordan T; Swanson, Chad M; Ahlquist, Paul; Sherer, Nathan M

    2016-04-01

    Retroviruses encode cis-acting RNA nuclear export elements that override nuclear retention of intron-containing viral mRNAs including the full-length, unspliced genomic RNAs (gRNAs) packaged into assembling virions. The HIV-1 Rev-response element (RRE) recruits the cellular nuclear export receptor CRM1 (also known as exportin-1/XPO1) using the viral protein Rev, while simple retroviruses encode constitutive transport elements (CTEs) that directly recruit components of the NXF1(Tap)/NXT1(p15) mRNA nuclear export machinery. How gRNA nuclear export is linked to trafficking machineries in the cytoplasm upstream of virus particle assembly is unknown. Here we used long-term (>24 h), multicolor live cell imaging to directly visualize HIV-1 gRNA nuclear export, translation, cytoplasmic trafficking, and virus particle production in single cells. We show that the HIV-1 RRE regulates unique, en masse, Rev- and CRM1-dependent "burst-like" transitions of mRNAs from the nucleus to flood the cytoplasm in a non-localized fashion. By contrast, the CTE derived from Mason-Pfizer monkey virus (M-PMV) links gRNAs to microtubules in the cytoplasm, driving them to cluster markedly to the centrosome that forms the pericentriolar core of the microtubule-organizing center (MTOC). Adding each export element to selected heterologous mRNAs was sufficient to confer each distinct export behavior, as was directing Rev/CRM1 or NXF1/NXT1 transport modules to mRNAs using a site-specific RNA tethering strategy. Moreover, multiple CTEs per transcript enhanced MTOC targeting, suggesting that a cooperative mechanism links NXF1/NXT1 to microtubules. Combined, these results reveal striking, unexpected features of retroviral gRNA nucleocytoplasmic transport and demonstrate roles for mRNA export elements that extend beyond nuclear pores to impact gRNA distribution in the cytoplasm. PMID:27070420

  2. HIV-1 and M-PMV RNA Nuclear Export Elements Program Viral Genomes for Distinct Cytoplasmic Trafficking Behaviors.

    PubMed

    Pocock, Ginger M; Becker, Jordan T; Swanson, Chad M; Ahlquist, Paul; Sherer, Nathan M

    2016-04-01

    Retroviruses encode cis-acting RNA nuclear export elements that override nuclear retention of intron-containing viral mRNAs including the full-length, unspliced genomic RNAs (gRNAs) packaged into assembling virions. The HIV-1 Rev-response element (RRE) recruits the cellular nuclear export receptor CRM1 (also known as exportin-1/XPO1) using the viral protein Rev, while simple retroviruses encode constitutive transport elements (CTEs) that directly recruit components of the NXF1(Tap)/NXT1(p15) mRNA nuclear export machinery. How gRNA nuclear export is linked to trafficking machineries in the cytoplasm upstream of virus particle assembly is unknown. Here we used long-term (>24 h), multicolor live cell imaging to directly visualize HIV-1 gRNA nuclear export, translation, cytoplasmic trafficking, and virus particle production in single cells. We show that the HIV-1 RRE regulates unique, en masse, Rev- and CRM1-dependent "burst-like" transitions of mRNAs from the nucleus to flood the cytoplasm in a non-localized fashion. By contrast, the CTE derived from Mason-Pfizer monkey virus (M-PMV) links gRNAs to microtubules in the cytoplasm, driving them to cluster markedly to the centrosome that forms the pericentriolar core of the microtubule-organizing center (MTOC). Adding each export element to selected heterologous mRNAs was sufficient to confer each distinct export behavior, as was directing Rev/CRM1 or NXF1/NXT1 transport modules to mRNAs using a site-specific RNA tethering strategy. Moreover, multiple CTEs per transcript enhanced MTOC targeting, suggesting that a cooperative mechanism links NXF1/NXT1 to microtubules. Combined, these results reveal striking, unexpected features of retroviral gRNA nucleocytoplasmic transport and demonstrate roles for mRNA export elements that extend beyond nuclear pores to impact gRNA distribution in the cytoplasm.

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

    SciTech Connect

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

    2009-10-14

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

  4. A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel

    NASA Astrophysics Data System (ADS)

    Lee, Young-Ho; Byun, Thak Sang

    2015-10-01

    Accident-tolerant fuels are expected to have considerably longer coping time to respond to the loss of active cooling under severe accidents and, at the same time, have comparable or improved fuel performance during normal operation. The wear resistance of accident tolerant fuels, therefore, needs to be examined to determine the applicability of these cladding candidates to the current operating PWRs because the most common failure of nuclear fuel claddings is still caused by grid-to-rod fretting during normal operations. In this study, reciprocating sliding wear tests on three kinds of cladding candidates for accident-tolerant fuels have been performed to investigate the tribological compatibilities of self-mated cladding candidates and to determine the direct applicability of conventional Zirconium-based alloys as supporting structural materials. The friction coefficients of the cladding candidates are strongly influenced by the test environments and coupled materials. The wear test results under water lubrication conditions indicate that the supporting structural materials for the cladding candidates of accident-tolerant fuels need to be replaced with the same cladding materials instead of using conventional Zirconium-based alloys.

  5. Prediction of Fracture Behavior in Rock and Rock-like Materials Using Discrete Element Models

    NASA Astrophysics Data System (ADS)

    Katsaga, T.; Young, P.

    2009-05-01

    The study of fracture initiation and propagation in heterogeneous materials such as rock and rock-like materials are of principal interest in the field of rock mechanics and rock engineering. It is crucial to study and investigate failure prediction and safety measures in civil and mining structures. Our work offers a practical approach to predict fracture behaviour using discrete element models. In this approach, the microstructures of materials are presented through the combination of clusters of bonded particles with different inter-cluster particle and bond properties, and intra-cluster bond properties. The geometry of clusters is transferred from information available from thin sections, computed tomography (CT) images and other visual presentation of the modeled material using customized AutoCAD built-in dialog- based Visual Basic Application. Exact microstructures of the tested sample, including fractures, faults, inclusions and void spaces can be duplicated in the discrete element models. Although the microstructural fabrics of rocks and rock-like structures may have different scale, fracture formation and propagation through these materials are alike and will follow similar mechanics. Synthetic material provides an excellent condition for validating the modelling approaches, as fracture behaviours are known with the well-defined composite's properties. Calibration of the macro-properties of matrix material and inclusions (aggregates), were followed with the overall mechanical material responses calibration by adjusting the interfacial properties. The discrete element model predicted similar fracture propagation features and path as that of the real sample material. The path of the fractures and matrix-inclusion interaction was compared using computed tomography images. Initiation and fracture formation in the model and real material were compared using Acoustic Emission data. Analysing the temporal and spatial evolution of AE events, collected during the

  6. Partitioning and Leaching Behavior of Actinides and Rare Earth Elements in a Zirconolite- Bearing Hydrothermal Vein System

    SciTech Connect

    Payne, Timothy E.; Hart, Kaye P.; Lumpkin, Gregory R.; McGlinn, Peter J.; Giere, Reto

    2007-07-01

    Chemical extraction techniques and scanning electron microscopy were used to study the distribution and behavior of actinides and rare earth elements (REE) in hydrothermal veins at Adamello (Italy). The six samples discussed in this paper were from the phlogopite zone, which is one of the major vein zones. The samples were similar in their bulk chemical composition, mineralogy, and leaching behavior of major elements (determined by extraction with 9 M HCl). However, there were major differences in the extractability of REE and actinides. The most significant influence on the leaching characteristics appears to be the amounts of U, Th and REE incorporated in resistant host phases (zirconolite and titanite) rather than readily leached phases (such as apatite). Uranium and Th are very highly enriched in zirconolite grains. Actinides were more readily leached from samples with a higher content of U and Th, relative to the amount of zirconium. The results show that REE and actinides present in chemically resistant host minerals can be retained under aggressive leaching conditions. (authors)

  7. Influence of alloying elements on the chlorination behavior of nickel- and iron-based alloys

    SciTech Connect

    Brill, U.; Kloewer, J.; Agarwal, D.C.

    1996-11-01

    A wide range of commercial heat-resistant alloys has been tested in a H{sub 2} + 10% HCl environment at 550 C, 650 C, 680 C, 750 C and 850 C. The tests were carried out using a 24 h cycle with a total test time of up to 1,056 H. Weight change was determined, and the average value for three specimens per alloy and temperature plotted versus time, followed by a metallographic examination of the depth of corrosion. By a statistical evaluation of the data generated, it was possible to describe the weight change and penetration depth of all the alloys under examination as a function of the concentration of their main alloying elements and test temperature. According to these results, alloying elements nickel and molybdenum have a beneficial influence on chlorination resistance, whereas silicon and titanium are detrimental. Increased temperature always resulted in enhance corrosion. Only Ni, Ni-Mo, and Ni-Cr-Mo alloys show acceptable resistance for temperatures up to 850 C.

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

    SciTech Connect

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

    1981-04-01

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

  9. Fretting wear behaviors of a dual-cooled nuclear fuel rod under a simulated rod vibration

    SciTech Connect

    Lee, Young-Ho; Kim, Hyung-Kyu; Kang, Heung-Seok; Yoon, Kyung-Ho; Kim, Jae-Yong; Lee, Kang-Hee

    2012-06-06

    Recently, a dual-cooled fuel (i.e., annular fuel) that is compatible with current operating PWR plants has been proposed in order to realize both a considerable amount of power uprating and an increase of safety margins. As the design concept should be compatible with current operating PWR plants, however, it shows a narrow gap between the fuel rods when compared with current solid nuclear fuel arrays and needs to modify the spacer grid shapes and their positions. In this study, fretting wear tests have been performed to evaluate the wear resistance of a dual-cooled fuel by using a proposed spring and dimple of spacer grids that have a cantilever type and hemispherical shape, respectively. As a result, the wear volume of the spring specimen gradually increases as the contact condition is changed from a certain gap, just contact to positive force. However, in the dimple specimen, just contact condition shows a large wear volume. In addition, a circular rod motion at upper region of contact surface is gradually increased and its diametric size depends on the wear depth increase. Based on the test results, the fretting wear resistance of the proposed spring and dimple is analyzed by comparing the wear measurement results and rod motion in detail.

  10. Fretting wear behaviors of a dual-cooled nuclear fuel rod under a simulated rod vibration

    NASA Astrophysics Data System (ADS)

    Lee, Young-Ho; Kim, Hyung-Kyu; Kang, Heung-Seok; Yoon, Kyung-Ho; Kim, Jae-Yong; Lee, Kang-Hee

    2012-06-01

    Recently, a dual-cooled fuel (i.e., annular fuel) that is compatible with current operating PWR plants has been proposed in order to realize both a considerable amount of power uprating and an increase of safety margins. As the design concept should be compatible with current operating PWR plants, however, it shows a narrow gap between the fuel rods when compared with current solid nuclear fuel arrays and needs to modify the spacer grid shapes and their positions. In this study, fretting wear tests have been performed to evaluate the wear resistance of a dual-cooled fuel by using a proposed spring and dimple of spacer grids that have a cantilever type and hemispherical shape, respectively. As a result, the wear volume of the spring specimen gradually increases as the contact condition is changed from a certain gap, just contact to positive force. However, in the dimple specimen, just contact condition shows a large wear volume. In addition, a circular rod motion at upper region of contact surface is gradually increased and its diametric size depends on the wear depth increase. Based on the test results, the fretting wear resistance of the proposed spring and dimple is analyzed by comparing the wear measurement results and rod motion in detail.

  11. Behavior of Rare Earth Element In Geothermal Systems; A New Exploration/Exploitation Tool

    SciTech Connect

    Scott A. Wood

    2002-01-28

    The goal of this four-year project was to provide a database by which to judge the utility of the rare earth elements (REE) in the exploration for and exploitation of geothermal fields in the United States. Geothermal fluids from hot springs and wells have been sampled from a number of locations, including: (1) the North Island of New Zealand (1 set of samples); (2) the Cascades of Oregon; (3) the Harney, Alvord Desert and Owyhee geothermal areas of Oregon; (4) the Dixie Valley and Beowawe fields in Nevada; (5) Palinpion, the Philippines: (6) the Salton Sea and Heber geothermal fields of southern California; and (7) the Dieng field in Central Java, Indonesia. We have analyzed the samples from all fields for REE except the last two.

  12. A finite element analysis of the freeze/thaw behavior of external artery heat pipes

    NASA Technical Reports Server (NTRS)

    Lu, X. J.; Peterson, G. P.

    1993-01-01

    A two-dimensional finite element model was used to determine the freeze/thaw characteristics of an external artery heat pipe. During startup, the working fluid, which was located in the liquid channel and the circumferential wall grooves, experienced a phase transformation from a solid to a liquid state. The transient heat conduction equations with moving interfacial conditions were solved using the appropriate initial boundary conditions. The modelling results include the cross-sectional temperature distribution and the interfacial or melt front position as a function of time. A fixed grid approach was adopted in the model for the phase-change process during thawing of frozen working fluid. The interfacial position between the liquid and solid regions was found by balancing the latent heat caused by interfacial movement with the heat addition or extraction at the related grid points.

  13. An experimental method for quantitatively evaluating the elemental processes of indoor radioactive aerosol behavior.

    PubMed

    Yamazawa, H; Yamada, S; Xu, Y; Hirao, S; Moriizumi, J

    2015-11-01

    An experimental method for quantitatively evaluating the elemental processes governing the indoor behaviour of naturally occurring radioactive aerosols was proposed. This method utilises transient response of aerosol concentrations to an artificial change in aerosol removal rate by turning on and off an air purifier. It was shown that the indoor-outdoor exchange rate and the indoor deposition rate could be estimated by a continuous measurement of outdoor and indoor aerosol number concentration measurements and by the method proposed in this study. Although the scatter of the estimated parameters is relatively large, both the methods gave consistent results. It was also found that the size distribution of radioactive aerosol particles and hence activity median aerodynamic diameter remained not largely affected by the operation of the air purifier, implying the predominance of the exchange and deposition processes over other processes causing change in the size distribution such as the size growth by coagulation and the size dependence of deposition.

  14. Effects of reactive element additions and sulfur removal on the oxidation behavior of FECRAL alloys

    SciTech Connect

    Stasik, M.C.; Pettit, F.S.; Meier, G.H. . Dept. of Materials Science and Engineering); Ashary, A. ); Smialek, J.L. )

    1994-12-15

    The results of this study have shown that desulfurization of FeCrAl alloys by hydrogen annealing can result in improvements in cyclic oxidation comparable to that achieved by doping with reactive elements. Moreover, specimens of substantial thicknesses can be effectively desulfurized because of the high diffusivity of sulfur in bcc iron alloys. The results have also shown that there is less stress generation during the cyclic oxidation of Y-doped FeCrAl compared to Ti-doped or desulfurized FeCrAl. This indicates that the growth mechanism, as well as the strength of the oxide/alloy interface, influences the ultimate oxidation morphology and stress state which will certainly affect the length of time the alumina remains protective.

  15. Mechanical behavior modeling of sand-rubber chips mixtures using discrete element method (DEM)

    NASA Astrophysics Data System (ADS)

    Eidgahee, Danial Rezazadeh; Hosseininia, Ehsan Seyedi

    2013-06-01

    Rubber shreds in mixture with sandy soils are widely used in geotechnical purposes due to their specific controlled compressibility characteristics and light weight. Various studies have been carried out for sand or rubber chips content in order to restrain the compressibility of the mass in different structures such as backfills, road embankments, etc. Considering different rubber contents, sand-rubber mixtures can be made which lead mechanical properties of the blend to go through changes. The aim of this paper is to study the effect of adding different rubber portions on the global engineering properties of the mixtures. This study is performed by using Discrete Element Method (DEM). The simulations showed that adding rubber up to a particular fraction can improve maximum bearing stress characteristics comparing to sand alone masses. Taking the difference between sand and rubber stiffness into account, the result interpretation can be developed to other soft and rigid particle mixtures such as powders or polymers.

  16. The Geochemistry of Technetium: A Summary of the Behavior of an Artificial Element in the Natural Environment

    SciTech Connect

    Icenhower, Jonathan P.; Qafoku, Nikolla; Martin, Wayne J.; Zachara, John M.

    2008-12-01

    Interest in the chemistry of technetium has only increased since its discovery in 1937, mainly because of the large and growing inventory of 99Tc generated during fission of 235U, its environmental mobility in oxidizing conditions, and its potential radiotoxicity. For every ton of enriched uranium fuel (3% 235U) that is consumed at a typical burn-up rate, nearly 1 kg of 99Tc is generated. Thus, the mass of 99Tc produced since 1993 has nearly quadrupled, and will likely to continue to increase if more emphasis is placed on nuclear power to slow the accumulation of atmospheric greenhouse gases. In order to gain a comprehensive understanding of the interaction of 99Tc and the natural environment, we review the sources of 99Tc in the nuclear fuel cycle, its chemical properties, radiochemistry, and biogeochemical behavior. We include an evaluation of the use of Re as a chemical analog of Tc, as well as a summary of the redox potential, thermodynamics, sorption, colloidal behavior, and interaction of humic substances with Tc, and the potential for re-oxidation and remobilization of Tc(IV). What emerges is a more complicated picture of Tc behavior than that of an easily tractable transition of Tc(VII) to Tc(IV) with consequent immobilization. Reducing conditions (+200 to +100 mV Eh) are generally thought necessary to cause reduction of Tc(VII) to Tc(IV), but far more important are the presence of reducing agents, such as Fe(II) sorbed onto mineral grains. Catalysis of Tc(VII) by surface-mediated Fe(II) will bring the mobile Tc(VII) species to a lower oxidation state and will form the relatively insoluble Tc(IV)O2∙nH2O, but even as a solid, equilibrium concentrations of aqueous Tc are nearly a factor of 20× above the EPA set drinking water standards. However, sequestration of Tc(IV) into Fe(III)-bearing phases, such as goethite or other hydrous oxyhydroxides of iron, may ameliorate concerns over the mobility of Tc. Further, the outcome of many studies on terrestrial and

  17. The Biogeochemistry of Technetium: A review of the behavior of an artificial element in the natural environment

    SciTech Connect

    Icenhower, Jonathan P.; Qafoku, Nikolla; Zachara, John M.; Martin, Wayne J.

    2010-10-04

    Interest in the chemistry of technetium has only increased since its discovery in 1937, mainly because of the large and growing inventory of 99Tc generated during fission of 235U, its environmental mobility in oxidizing conditions, and its potential radiotoxicity. For every ton of enriched uranium fuel (3% 235U) that is consumed at a typical burn-up rate, nearly 1 kg of 99Tc is generated. Thus, the mass of 99Tc produced since 1993 has nearly quadrupled, and the pace of generation will likely increase if more emphasis is placed on nuclear power to slow the accumulation of atmospheric greenhouse gases. In order to gain a comprehensive understanding of the interaction of 99Tc and the natural environment, we review the sources of 99Tc in the nuclear fuel cycle and its biogeochemical behavior. We include an evaluation of the use of Re as a chemical analog of Tc, as well as a summary of the redox potential, sorption, colloidal behavior, and interaction of humic substances with Tc, and the potential for re-oxidation and remobilization of Tc(IV). What emerges is a more complicated picture of Tc behavior than that of an easily tractable transition of Tc(VII) to Tc(IV) with consequent immobilization. Reducing conditions (+200 to +100 mV Eh) and the presence of Fe(II) sorbed onto Fe(III) (oxy)hydroxides will bring the mobile Tc(VII) species to a lower oxidation state and will form the relatively insoluble Tc(IV)O2∙nH2O, but even as a solid, equilibrium concentrations of aqueous Tc are nearly a factor of 20× above the EPA set drinking water standards. However, sequestration of Tc(IV) into Fe(III)-bearing phases, such as goethite, iron-bearing phyllosilicates and, perhaps, siderite, may ameliorate concerns over the mobility of Tc. A key factor, elucidated through experiment, in retarding the mobility of Tc in the environment is isolation from exposure to oxygen, which occurs when Tc is in a crystallographic position in a solid phase.

  18. Distribution of toxic elements in the products of the extraction of bitumens from coal and peat

    SciTech Connect

    S.I. Zherebtsov; M.Yu. Klimovich; A.I. Moiseev

    2008-06-15

    The effect of bitumen extraction conditions on the behavior of a number of toxic trace elements (V, Cr, Ni, Cu, Zn, As, Sb, and Pb) in solid fossil fuel samples was studied. It was found that the capacity of fossil fuels to concentrate trace elements can be monitored and the type of compound can be determined based on the dimensionless quantities of carbophilicity and concentration factors of the trace elements in coal and peat. The types of compounds in fossil fuels were found and graphically represented for the test trace elements.

  19. Gas-leaking fuel elements number and fission gas product coolant volumetric activities assessment in the VVER-440 nuclear power plant

    NASA Astrophysics Data System (ADS)

    Szuta, Marcin

    1992-07-01

    In a nuclear power plant it is required to monitor continuously the number of gas-leaking fuel elements and the contamination level of the primary coolant by fission gas products. It is proposed to use the radiation monitoring system equipped with the computer technics provided with the suitable program package for fulfilment this requirements. The input data to start up the program consists of the 88Kr volumetric activity measured by the radiation monitoring system and three actual technological parameters: coolant temperature at inlet, thermal power and coolant flow rate.

  20. The Behavior of Rare Earth and Other Trace Elements During Laboratory Melting of the Mantle at 1.0 GPa.

    NASA Astrophysics Data System (ADS)

    Johnston, A.; Schwab, B. E.; Witter, J. P.

    2002-12-01

    Earlier piston-cylinder experiments in our laboratory produced a collection of mantle melting run products that have now been analyzed by ion probe for selected REE, Ti, Cr, Rb, Sr, Y, Zr, and Nb. Starting materials consisted of five fertile to intermediate, lherzolitic to wehrlitic mixtures of natural ol, cpx, opx, and sp handpicked from fresh xenoliths. Samples were run in graphite-lined Pt capsules and the melt was separated from the residual minerals into a layer of vitreous carbon spheres (VCS) thus circumventing the problems of Fe-loss and quench modification of the melt. Major element compositions of all phases were determined previously by electron microprobe and least-squares inversion of these data yielded modes for all run products. The bulk starting materials were analyzed for trace and major elements by ICP-MS and-ES at Boston University. The principle goals of the study were to evaluate whether the trace element data support the conclusion reached previously from the major element data that these run products represent very close approaches to equilibrium, and to evaluate whether the glass data set could be inverted to yield meaningful mineral/melt kd's. With few exceptions, we were unable to get good data from the crystalline phases, primarily because of their small sizes or very low trace element abundances. However, the glass phase in 32 run products (representing F's from ~2-50 wt. percent) yielded excellent data that were remarkably homogenous from spot to spot and varied sensibly with changing melt fraction. Forward modeling using our modes and Co values in conjunction with published kd's for ol, cpx, opx, and sp (Kelemen et. al. EPSL. 120: 111-134, 1993) yield calculated trace element abundances that generally agree with our measurements to within 10-30 percent, about the precision of the ion probe measurements, given the small beam diameter we employed. However, our attempts to run the inverse problem using our measurements, modes, and Co