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Sample records for fuel element behavior

  1. 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.

  2. 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)

  3. 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)

  4. Comparison of irradiation behavior of different uranium silicide dispersion fuel element designs

    SciTech Connect

    Hofman, G.L.; Rest, J.; Snelgrove, J.L.

    1995-01-01

    Calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2} were performed for various dispersion fuel element designs. Breakaway swelling criteria in the form of critical fuel volume fractions were derived with data obtained from U{sub 3}SiAl-Al plate irradiations. The results of the analysis show that rod-type elements remain well below the pillowing threshold. However, tubular fuel elements, which behave essentially like plates, will likely develop pillows or blisters at around 90% {sup 235}U burnup. The U{sub 3}Si{sub 2}-Al compounds demonstrate stable swelling behavior throughout the entire burnup range for all fuel element designs.

  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. Review of behavior of mixed-oxide fuel elements in extended overpower transient tests in EBR-II

    SciTech Connect

    Tsai, H.; Neimark, L.A.; Nagai, S.; Nakae, N.

    1994-10-01

    From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with varying design and prior operating histories. Elements with aggressive design features, such as high fuel smear density and/or thin cladding, were included to accentuate transient effects. The ramp rates were either 0.1 or 10% {Delta}P/P/s and the overpowers ranged between {approx}60 and 100% of the elements` prior power ratings. Six elements breached during the tests, all with aggressive design parameters. The other elements, including all those with moderate design features for the reference or advanced long-life drivers for PNC`s prototype fast reactor Monju, maintained their cladding integrity during the tests. Posttest examination results indicated that fuel/cladding mechanical interaction (FCMI) was the most significant mechanism causing the cladding strain and breach. In contrast, pressure loading from the fission gas in the element plenum was less important, even in high-burnup elements. During an overpower transient, FCMI arises from fuel/cladding differential thermal expansion, transient fuel swelling, and, significantly, the gas pressure in the sealed central cavity of elements with substantial centerline fuel melting. Fuel performance data from these tests, including cladding breaching margin and transient cladding strain, are correlatable with fuel-element design and operating parameters. These correlations are being incorporated into fuel-element behavior codes. At the two tested ramp rates, fuel element behavior appears to be insensitive to transient ramp rate and there appears to be no particular vulnerability to slow ramp transients as previously perceived.

  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. 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)

  9. 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

  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. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

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

    1962-04-10

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

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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.

  17. 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)

  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 INTERLOCKING ARRANGEMENT

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1963-01-01

    This patent relates to a system for mutually interlocking a multiplicity of elongated, parallel, coextensive, upright reactor fuel elements so as to render a laterally selfsupporting bundle, while admitting of concurrent, selective, vertical withdrawal of a sizeable number of elements without any of the remaining elements toppling, Each element is provided with a generally rectangular end cap. When a rank of caps is aligned in square contact, each free edge centrally defines an outwardly profecting dovetail, and extremitally cooperates with its adjacent cap by defining a juxtaposed half of a dovetail- receptive mortise. Successive ranks are staggered to afford mating of their dovetails and mortises. (AEC)

  20. Experimental evaluation of thermal ratcheting behavior in UO2 fuel elements

    NASA Technical Reports Server (NTRS)

    Phillips, W. M.

    1973-01-01

    The effects of thermal cycling of UO2 at high temperatures has been experimentally evaluated to determine the rates of distortion of UO2/clad fuel elements. Two capsules were rested in the 1500 C range, one with a 50 C thermal cycle, the other with a 100 C thermal cycle. It was observed that eight hours at the lower cycle temperature produced sufficient UO2 redistribution to cause clad distortion. The amount of distortion produced by the 100 C cycle was less than double that produced by the 50 C, indicating smaller thermal cycles would result in clad distortion. An incubation period was observed to occur before the onset of distortion with cycling similar to fuel swelling observed in-pile at these temperatures.

  1. 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.

  2. 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.

  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.

    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.

  7. 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.

  8. 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.

  9. 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)

  10. 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

  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. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Dickson, J.J.

    1963-09-24

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

  13. 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)

  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. 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)

  16. 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.

  17. 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.

  18. 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)

  19. 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)

  20. FUEL ELEMENT FOR NEUTRONIC REACTORS

    DOEpatents

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

    1961-01-17

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

  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. 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.

  3. 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)

  4. 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.

  5. 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.

  6. 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.

  7. Partitioning behavior of trace elements during pilot-scale combustion of pulverized coal and coal-water slurry fuel

    PubMed

    Nodelman; Pisupati; Miller; Scaroni

    2000-05-29

    Release pathways for inorganic hazardous air pollutants (IHAPs) from a pilot-scale, down-fired combustor (DFC) when firing pulverized coal (PC) and coal-water slurry fuel (CWSF) were identified and quantified to demonstrate the effect of fuel form on IHAP partitioning, enrichment and emissions. The baghouse capturing efficiency for each element was calculated to determine the effectiveness of IHAP emission control. Most of the IHAPs were enriched in the fly ash and depleted in the bottom ash. Mercury was found to be enriched in the flue gas, and preferentially emitted in the vapor phase. When firing CWSF, more IHAPs were partitioned in the bottom ash than when firing PC. Significant reduction of Hg emissions during CWSF combustion was also observed.

  8. 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.

  9. Thermionic fuel element technology status

    NASA Technical Reports Server (NTRS)

    Holland, J. W.; Horner, M. W.; Yang, L.

    1985-01-01

    The results of research, conducted between the mid-1960s and 1973, on the multiconverter thermionic fuel elements (TFEs) that comprise the reactor core of an SP-100 thermionic reactor system are presented. Fueled-emitter technology, insulator technology and cell and TFE assembly technology of the prototypical TFEs which were tested in-pile and out-of-pile during these years are described. The proto-TFEs have demonstrated reproducible performance within 5 percent and no premature failures within the 1.5 yr of operation (with projected 3-yr lifetimes). The two primary life-limiting factors had been identified as thermionic emitter dimensional increase due to interactions with the fuel and electrical insulator structural damage from fast neutrons. Multiple options for extending TFE lifetimes to 7 yr or longer are available and will be investigated in the 1984-1985 SP-100 program for resolution of critical technology issues. Design diagrams and test graphs are included.

  10. Fuel elements of research reactors in China

    SciTech Connect

    Yongmao, Z.; Dianshan, C.; Guofang, Q.

    1988-01-01

    This paper describes the current status of design, fabrication of fuel elements for research reactors in China, emphasis is placed on the technology of fuel elements for the High Flux Engineering Test Reactor (HFETR).

  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. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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

  17. 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)

  18. 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.

  19. MRT fuel element inspection at Dounreay

    SciTech Connect

    Gibson, J.

    1997-08-01

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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Behavioral Health Program Element

    NASA Technical Reports Server (NTRS)

    Leveton, Lauren B.

    2006-01-01

    The project goal is to develop behavioral health prevention and maintenance system for continued crew health, safety, and performance for exploration missions. The basic scope includes a) Operationally-relevant research related to clinical cognitive and behavioral health of crewmembers; b) Ground-based studies using analog environments (Antarctic, NEEMO, simulations, and other testbeds; c) ISS studies (ISSMP) focusing on operational issues related to behavioral health outcomes and standards; d) Technology development activities for monitoring and diagnostic tools; and e) Cross-disciplinary research (e.g., human factors and habitability research, skeletal muscle, radiation).

  6. IMPROVED TYPE OF FUEL ELEMENT

    DOEpatents

    Monson, H.O.

    1961-01-24

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

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Postirradiation examination of thermionic fuel element specimens

    SciTech Connect

    Cannon, N.S.; Lawrence, L.A.; Veca, A.R.

    1988-12-01

    The Thermionic Fuel Element (TFE) Verification Program is funded by the Department of Energy (DOE) with the objective of demonstrating a fuel element design for a multimegawatt-class thermionic reactor for space power systems (Bohl and Ranken 1987). A number of contractors and DOE laboratories are involved in this program. These include General Atomics (GA), which is responsible for the overall technical development, fabrication, and processing of components and TFE prototypes for fast reactor testing and Westinghouse Hanford Company (WHC), which has the responsibility for implementation of the fast reactor irradiation program. 1 ref., 3 figs.

  13. 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.

  14. Liquid fuel injection elements for rocket engines

    NASA Technical Reports Server (NTRS)

    Cox, George B., Jr. (Inventor)

    1993-01-01

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

  15. Stochastic modelling of power reactor fuel behavior

    NASA Astrophysics Data System (ADS)

    Mirza, Shahid Nawaz

    An understanding of the in-reactor behavior of nuclear fuel is essential to the safe and economic operation of a nuclear power plant. It is no longer possible to achieve this without computer code calculations. A state of art computer code, FRODO, for Fuel ROD Operation, has been developed to model the steady state behavior of fuel pins in a light water reactor and to do sensitivity analysis. FRODO concentrates on the thermal performance, fission product release and pellet-clad interaction and can be used to predict the fuel failure under the prevailing conditions. FRODO incorporates the numerous uncertainties involved in fuel behavior modeling, using statistical methods, to ascertain fuel failures and their causes. Sensitivity of fuel failure to different fuel parameters and reactor conditions can be easily evaluated. FRODO has been used to analyze the sensitivities of fuel failures to coolant flow reductions. It is found that the uncertainties have pronounced effects on conclusions about fuel failures and their causes.

  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. 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)

  18. CONCENTRIC TUBE FUEL ELEMENT SPRING ALIGNMENT SPACER DEVICE

    DOEpatents

    Weems, S.J.

    1963-09-24

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

  19. 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)

  20. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    SciTech Connect

    Montierth, Leland M.

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  1. 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)

  2. 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.

  3. Drying results of K-Basin fuel element 0309M (Run 3)

    SciTech Connect

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

    1998-07-01

    An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-West Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from K-West canister 0309M during the second fuel selection campaign, conducted in 1996, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. The fuel element was broken in two pieces, with a relatively clean fracture, and the larger piece was tested. A gray/white coating was observed. This was the first test of a damaged fuel element in the furnace. K-West canisters can hold up to seven complete fuel assemblies, but, for purposes of this report, the element tested here is designated as Element 0309M. Element 0309M was subjected to drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step.

  4. 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.

  5. Fuel Element Transfer Cask Modelling Using MCNP Technique

    SciTech Connect

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-05

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

  6. Fuel Element Transfer Cask Modelling Using MCNP Technique

    NASA Astrophysics Data System (ADS)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

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

  7. 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

  8. Postirradiation examinations of U-Pu-Zr fuel elements from subassemblies X419 and X419A

    SciTech Connect

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

    1986-10-01

    Initial postirradiation examination of IFR type U-Pu-Zr fuel elements from X419 and X419A are reported. Characterization of the fuel at three levels of burnup, 0.8 at.%, 1.9 at.%, and 2.7 at.% is presented. Fuel swelling, microstructure, chemical redistribution, and fission gas behavior is discussed. No evidence was found for any performance-limiting damage to the fuel elements at these burnups.

  9. 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.

  10. 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.

  11. Assay method for MTR-type fuel elements

    SciTech Connect

    Sher, R.; Shea, P.

    1981-01-01

    This paper describes a calculation procedure that can be used by IAEA inspectors to verify unirradiated MTR-type fuel elements. The procedure is programmable on a small programmable calculator (HP-97). The accuracy of the calculation enables the inspector to determine whether the element contains the correct number of fuel plates of the stated design. 2 refs.

  12. 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.

  13. Hydride fuel behavior in LWRs

    NASA Astrophysics Data System (ADS)

    Olander, Donald R.; Ng, Marowen

    2005-11-01

    The U-Zr hydride U 0.31ZrH 1.6 offers a number of advantages over oxide fuel for light-water reactors. Fission-gas release appears to be very small (release fraction ˜10 -4) up to 600 °C, which is close to the maximum fuel temperature. Initial irradiation-induced swelling can be as large as 5% for temperatures exceeding 650 °C. Hydrogen redistributes due to the non-uniform temperature in the fuel from the as-fabricated H/Zr of 1.6 to one that is higher at the pellet periphery than at the centerline. Radial redistribution produces 'hydrogen' stresses in the pellet which add to the usual thermal stresses. In a helium-bonded fuel rod, the total stresses are less than the fracture stress; in a liquid-metal-bonded fuel rod, the fracture stress is exceeded in the central portion of the pellet, but the surface remains in compression. Axial redistribution moves substantial quantities of hydrogen from the middle portion of the fuel stack to the ends. The neutronic effect of this displacement of the moderator is unknown.

  14. Nuclear fuel elements and method of making same

    DOEpatents

    Schweitzer, Donald G.

    1992-01-01

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

  15. 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.

  16. 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.

  17. 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.

  18. Technology Status of Thermionic Fuel Elements for Space Nuclear Power

    NASA Technical Reports Server (NTRS)

    Holland, J. W.; Yang, L.

    1984-01-01

    Thermionic reactor power systems are discussed with respect to their suitability for space missions. The technology status of thermionic emitters and sheath insulator assemblies is described along with testing of the thermionic fuel elements.

  19. 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.

  20. 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

  1. 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

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

    NASA Astrophysics Data System (ADS)

    Higgs, Jamie D.

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

  3. 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.

  4. Spent fuel behavior in dry storage

    NASA Astrophysics Data System (ADS)

    Johnson, A. B., Jr.; Pankaskie, P. J.; Gilbert, E. R.

    1982-02-01

    Dry storage is emerging as an attractive and timely alternative to complement wet storage, and assist utilities to meet interim storage needs. Spent fuel is handled and stored under dry conditions. Dry storage of irradiated Zircaloy clad fuel in metal casks, drywells, silos and vaults is demonstrated. Hot cell and laboratory studies also are underway to investigate specific phenomena related to cladding behavior in dry storage. A substantial fraction of the LWR spent fuel inventory has aged for relatively long times and has relatively low decay heats. This suggests that much of the fuel inventory can be stored at relatively low temperatures. Alternatively, rod consolidation of the older can be considered without exceeding maximum cladding temperatures.

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

    SciTech Connect

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

    2016-01-22

    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 ({sup 233}U), 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.

  6. 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.

  7. 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)

  8. 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.

  9. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

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

    1993-01-14

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

  10. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

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

    1993-05-01

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

  11. Dart model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

    SciTech Connect

    Rest, J.; Hofman, G.L.

    1997-06-01

    The Dispersion Analysis Research Tool (DART) contains models for fission-gas induced fuel swelling, interaction of fuel with the matrix aluminum, resultant reaction-product swelling, and calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data. DART results are compared with data for fuel swelling Of U{sub 3}SiAl-Al in plate, tube, and rod configurations as a function of fission density. Plate and tube calculations were performed at a constant fuel temperature of 373 K and 518 K, respectively. An irradiation temperature of 518 K results in a calculated aluminide layer thickness for the Russian tube that is in the center of the measured range (16 {mu}m). Rod calculations were performed with a temperature gradient across the rod characterized by surface and central temperatures of 373 K and 423 K, respectively. The effective yield stress of irradiated Al matrix material and the aluminide was determined by comparing the results of DART calculations with postirradiation immersion volume measurement of U{sub 3}SiAl plates. The values for the effective yield stress were used in all subsequent simulations. The lower calculated fuel swelling in the rod-type element is due to an assumed biaxial stress state. Fuel swelling in plates results in plate thickness increase only. Likewise, in tubes, only the wall thickness increases. Irradiation experiments have shown that plate-type dispersion fuel elements can develop blisters or pillows at high U-235 burnup when fuel compounds exhibiting breakaway swelling are used at moderate to high fuel volume fractions. DART-calculated interaction layer thickness and fuel swelling follows the trends of the observations. 3 refs., 2 figs.

  12. 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.

  13. 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.

  14. Methods for making a porous nuclear fuel element

    DOEpatents

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

    2014-12-30

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

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

    NASA Astrophysics Data System (ADS)

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

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

  16. 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.

  17. 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}.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

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

    SciTech Connect

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

    2001-10-01

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

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

    NASA Technical Reports Server (NTRS)

    Gomez, C. F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.

    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.

  5. 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)

  6. 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.

  7. 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

  8. 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)

  9. 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)

  10. 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.

  11. 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.

    2016-11-01

    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.

  12. 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.

  13. 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.

  14. FISSILE MATERIAL AND FUEL ELEMENTS FOR NEUTRONIC REACTORS

    DOEpatents

    Shaner, B.E.

    1961-08-15

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

  15. 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.

  16. Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

    NASA Astrophysics Data System (ADS)

    Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

    2017-04-01

    Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

  17. Fuel cladding behavior under rapid loading conditions

    NASA Astrophysics Data System (ADS)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  18. 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.

  19. 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.

  20. 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.

  1. 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)

  2. 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)

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

    DOEpatents

    Handwerk, J.H.; BAch, R.A.

    1959-08-18

    A method is described for preparing a reactor fuel element by forming a mixture of thorium dioxide and an oxide of uranium, the uranium being present. In an oxidation state at least as high as it is in U/sub 3/O/sub 8/, into a desired shape and firing in air at a temperature siifficiently high to reduce the higher uranium oxide to uranium dioxide.

  4. 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

  5. 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.

  6. 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.

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

    SciTech Connect

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

    1998-07-01

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

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

    SciTech Connect

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

    1998-05-01

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

  9. A novel microbial fuel cell sensor with biocathode sensing element.

    PubMed

    Jiang, Yong; Liang, Peng; Liu, Panpan; Wang, Donglin; Miao, Bo; Huang, Xia

    2017-03-02

    The traditional microbial fuel cell (MFC) sensor with bioanode as sensing element delivers limited sensitivity to toxicity monitoring, restricted application to only anaerobic and organic rich water body, and increased potential fault warning to the combined shock of organic matter/toxicity. In this study, the biocathode for oxygen reduction reaction was employed for the first time as the sensing element in MFC sensor for toxicity monitoring. The results shown that the sensitivity of MFC sensor with biocathode sensing element (7.4±2.0 to 67.5±4.0mA%(-1)cm(-2)) was much greater than that showed by bioanode sensing element (3.4±1.5 to 5.5±0.7mA%(-1)cm(-2)). The biocathode sensing element achieved the lowest detection limit reported to date using MFC sensor for formaldehyde detection (0.0005%), while the bioanode was more applicable for higher concentration (>0.0025%). There was a quicker response of biocathode sensing element with the increase of conductivity and dissolved oxygen (DO). The biocathode sensing element made the MFC sensor directly applied to clean water body monitoring, e.g., drinking water and reclaimed water, without the amending of background organic matter, and it also decreased the warning failure when challenged by a combined shock of organic matter/toxicity.

  10. 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.

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

    SciTech Connect

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

    2013-01-01

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

  12. Irradiation behavior of metallic fast reactor fuels

    SciTech Connect

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985.

  13. An assessment of swaged connections in a nuclear fuel element using nonlinear finite element analysis

    SciTech Connect

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

    1995-12-01

    Large displacement, non-linear finite element analyses were performed to evaluate a swaging process used to fabricate connections between plates in the fuel elements for a test reactor at the Idaho National Engineering Laboratory. The force required to pull the fuel plate from the connection is referred to as the strength of the connection. Assurance that the integrity of the connections is maintained through reactor operation is provided by establishing a minimum acceptance requirement for this strength. Analysis results were used to assess the sensitivity of the strength of the swaged connections to variations in several manufacturing process parameters. The predicted strengths correlated well with results from tests where sample swaged connections were loaded to failure. Results from these investigations were used to assess the adequacy and need for various fabrication, testing, and quality control requirements.

  14. Solid oxide fuel cell stacks using extruded honeycomb type elements

    NASA Astrophysics Data System (ADS)

    Wetzko, M.; Belzner, A.; Rohr, F. J.; Harbach, F.

    A solid oxide fuel cell (SOFC) stack concept is described which comprises "condensed-tubes" like extruded honeycomb sections of ceramic electrolyte (ZrO 2-based) and interconnectors of nickel sheet as key elements. According to this concept, well known and extensively tested construction principles can be realised in a low-cost production. The cells are self-supported with in-plane conduction. A demonstrator model stack of five honeycomb elements and six nickel sheet seals/interconnectors was built and operated for 860 h at 1000°C. Volumetric power densities of 160 kW/m 3 were obtained with H 2 vs. air, of close to 200 kW/m 3 with H 2 vs. O 2.

  15. Triaxial Swirl Injector Element for Liquid-Fueled Engines

    NASA Technical Reports Server (NTRS)

    Muss, Jeff

    2010-01-01

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

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

    SciTech Connect

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

    1982-04-01

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

  17. Correlation between annealing and irradiation behavior of dispersion fuels

    SciTech Connect

    Wiencek, T.C.; Domagala, R.F.

    1987-01-01

    Studying the effects of annealing of scaled-down aluminum matrix dispersion fuel plates is an important part of the data base for fuel performance. One of the most critical aspects of fuel performance is the stability of a fuel/matrix dispersion, which is usually measured by volumetric changes of the fuel zone. The preferred response of any fuel under irradiation would be no change in volume or a small, steady predictable change. These volume changes decrease the cooling gap channels and restrict the useful life of an element. Previous studies have defined the degree of volumetric change in current high-loading uranium silicide test reactor fuels. It has been demonstrated that some fuels behave well under irradiation, while others go into a breakaway swelling mode. A correlation has been proposed that fission-induced amorphization is responsible for the instability of the fuel and that such transformations can be predicted by the thermodynamic properties of the fuel. To complement this theory, it is proposed that annealing studies may be used as a screening test for new fuels for which no thermodynamic properties have been measured and/or no irradiation data are available. Irradiation performance could be estimated faster and without the expense of irradiating the fuels under investigation.

  18. 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.

  19. Steady-state fuel behavior modeling of nitride fuels in FRAPCON-EP

    NASA Astrophysics Data System (ADS)

    Feng, Bo; Karahan, Aydın; Kazimi, Mujid S.

    2012-08-01

    Fuel material properties and mechanistic fission gas models in FRAPCON-EP were updated to model the steady-state behavior of high-porosity nitride fuel operating at temperatures below half of the melting point. The fuel thermal conductivity and fuel thermal expansion models were updated with correlations for UN and (U,Pu)N fuels. Hot-pressing of the as-fabricated porosity was modeled as a function of the hydrostatic pressure and creep rate. The solid fission product swelling was assumed to increase linearly with burnup. Fission gas swelling constitutive models were updated to appropriately capture the intragranular gas bubble evolution in nitride fuel. Intergranular gas swelling was neglected due to the assumed high porosity of the fuel. The fission gas release behavior was modeled by fitting the fission gas diffusion coefficient in UN to FRAPCON's default fission gas release model. This fitted gas diffusion coefficient reflects the effects of porosity, burnup, operating temperature, fission rate, and bubble sink strength. Fission gas release and fuel swelling benchmarks against irradiation data were performed. The updated code was applied to UN fuel in typical PWR geometry and operating conditions, with an extended cycle length of 24 months. The results show that swelling of the nitride fuel up to 60 MWd/kg burnup did not lead to excessive straining of the cladding. Furthermore, this study showed that a porous (>15% porosity) nitride fuel pellet could achieve a much higher margin to failure from the cladding collapse and grid-to-rod fretting.

  20. 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.

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

    SciTech Connect

    Freels, James D; Jain, Prashant K

    2011-01-01

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

  2. 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.

  3. Strategy for phase 2 whole element furnace testing K West fuel

    SciTech Connect

    Lawrence, L.A.

    1998-03-13

    A strategy was developed for the second phase of the whole element furnace testing of damaged fuel removed from the K West Basin. The Phase 2 testing can be divided into three groups covering oxidation of whole element in moist inert atmospheres, drying elements for post Cold Vacuum Drying staging tests, and drying additional K West elements to provide confirmation of the results from the first series of damaged K West fuel drying studies.

  4. 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.

  5. 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

  6. Spherical fuel elements for advanced HTR manufacture and qualification by irradiation testing

    NASA Astrophysics Data System (ADS)

    Mehner, A.-W.; Heit, W.; Röllig, K.; Ragoss, H.; Müller, H.

    1990-04-01

    The reference fuel cycle for future pebble bed HTRs uses low enriched uranium fuel. The spherical fuel element for these HTRs is a 60 mm diameter sphere containing TRISO-coated particles with UO 2 kernels. Qualification of this fuel was performed by production and quality control experience, irradiation testing and accident simulation experiments. The results of the qualification programme fully support the new safety concepts of advanced HTR designs. Further work concentrates on consolidating performance data sets and on quantifying the endurance limits of reference fuel elements under normal and accident conditions.

  7. 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.

  8. 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 ...

  9. 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.

  10. 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).

  11. Drying results of K-Basin fuel element 5744U (Run 4)

    SciTech Connect

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

    1998-07-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fourth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 5744U. This element (referred to as Element 5744U) was stored underwater in the K-West Basin from 1983 until 1996. Element 5744U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0.

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

    SciTech Connect

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

    1999-07-26

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

  13. Drying Results of K-Basin Fuel Element 6513U (Run 8)

    SciTech Connect

    BM Oliver; GS Klinger; J Abrefah; SC Marschman; PJ MacFarlan; GA Ritter

    1999-08-11

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the eighth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 6513U. This element (referred to as Element 6513U) was stored underwater in the K-West Basin from 1983 until 1996. Element 6513U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0 and discussed in Section 6.0.

  14. Drying results of K-Basin fuel element 1164M (run 6)

    SciTech Connect

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

    1998-08-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the sixth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 1164 M. This element (referred to as Element 1164M) was stored underwater in the K-West Basin from 1983 until 1996. Element 1164M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0.

  15. 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.

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

    SciTech Connect

    Konyashov, Vadim V.; Krasnov, Alexander M.

    2002-04-15

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

  17. 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.

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

    SciTech Connect

    Gay, R.L.

    1995-09-12

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

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

    SciTech Connect

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

    1982-01-01

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

  20. 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…

  1. Genuineness: An Overlooked Element of Inviting Behavior

    ERIC Educational Resources Information Center

    Shaw, Daniel E.

    2004-01-01

    Genuineness, the condition of sincerity in behaving as one's unadulterated self, is a concept that is inherently congruent with inviting behavior yet is not directly addressed in the literature concerned with Invitational Theory. The author discusses the concept of genuineness, describes how it fits Invitational Theory and argues for its inclusion…

  2. 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.

  3. 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.

  4. 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.

  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, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Fuel Element Fabrication Plant... Appendix O to Part 110—Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority Note: Nuclear fuel elements are manufactured from source or...

  6. 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.

  7. 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.

  8. 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

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

    NASA Technical Reports Server (NTRS)

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

    1990-01-01

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

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

    PubMed

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

    1985-01-01

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

  11. 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)

  12. Mathematical simulation of hydrocarbon fuel conversion in heat-protection elements of hypersonic aircrafts

    NASA Astrophysics Data System (ADS)

    Kuranov, A. L.; Korabel'nikov, A. V.; Mikhailov, A. M.

    2017-01-01

    We consider a mathematical model of hydrocarbon fuel conversion in a thermochemical reactor as an element of heat protection of a hypersonic aircraft. The application of this model has made it possible to enrich information obtained in experimental studies.

  13. 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.

  14. Drying results of K-Basin fuel element 1990 (Run 1)

    SciTech Connect

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

    1998-06-01

    The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0.

  15. Accuracy of trace element determinations in alternate fuels

    NASA Technical Reports Server (NTRS)

    Greenbauer-Seng, L. A.

    1980-01-01

    NASA-Lewis Research Center's work on accurate measurement of trace level of metals in various fuels is presented. The differences between laboratories and between analytical techniques especially for concentrations below 10 ppm, are discussed, detailing the Atomic Absorption Spectrometry (AAS) and DC Arc Emission Spectrometry (dc arc) techniques used by NASA-Lewis. Also presented is the design of an Interlaboratory Study which is considering the following factors: laboratory, analytical technique, fuel type, concentration and ashing additive.

  16. Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

    SciTech Connect

    Miao, Yinbin; Mo, Kun; Yacout, Abdellatif; Harp, Jason

    2016-09-15

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U3Si2 as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U3Si2 fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U3Si2. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

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

    NASA Technical Reports Server (NTRS)

    Vezhba, I.

    1979-01-01

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

  18. Analysis of Geochemical Element Behaviors in Soils by Multifractal Methods

    NASA Astrophysics Data System (ADS)

    Xie, S.; Cheng, Q.; Huang, K.; Bao, Z.

    2009-04-01

    Soils are not only one of crucial environmental media but also one of the most important objects for environmental geochemistry. Usually, soils are developed from various kinds of rocks. During the weathering of rocks, the issues, such as the evolution of geochemical elements and the different behaviors of geochemical elements in soils, have been gaining great concern. This paper will explore the geochemical distribution patterns by fractal and multifractal methods, and then use the soils with different periods in the Hainan Island, China, as the case study to analyze the behaviors of geochemical elements. The soils in the Hainan Island are strongly aluminized, and anthropological activities have little impacts on the zonal soils. Thus the dynamic evolutions of the geochemical elements in soils are relatively of their original characteristics. The TAS analysis based on the diagram of SiO2-Na2O+K2O is implemented and has shown that most of the soils are evolved from granodiorite rocks, and some from basalts. This analytical result is quite consistent to the regional rock sampling investigation. Spider diagrams of trace elements also present that the soils are mainly developed from basalts and granites. In terms of the multifractal parameters characterizing the distribution patterns of the geochemical elements in the soils evolved from seven geological systems, it is shown that in the surroundings the multifractal behaviors of the geochemical elements are strongly associated with their inner geochemical behaviors on one hand, and on the other hand, anthropological activities and weathering behaviors of the soils also have specific effects on the multifractal parameters. In the study, the right parts of the multifractal spectrum curves of the geochemical elements in soils evolved from the Changchengian system change significantly, which implies that there are much lower geochemical data for elements. But for the soils evolved from granites and basalts, the left parts

  19. 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.

  20. 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.

  1. 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.

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

    SciTech Connect

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

    1999-09-24

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

  3. Test plan for surface and subsurface examinations of K-east and K-west fuel elements

    SciTech Connect

    Pitner, A.L.

    1997-04-14

    The test plan for subsurface examinations on damaged K East and K West Basin fuel elements is presented. The purpose of these examinations is to inspect damaged areas on the fuel elements for the presence of voids, sludge, or broken fuel, and to obtain samples from the damaged areas for subsequent characterization tests.

  4. 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.

  5. 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%.

  6. Leaker B.W.R. spent fuel elements: Radiochemical analysis on cover gases of storage containers after long storage

    SciTech Connect

    Paratore, A.L.; Pastore, G.; Partiti, C.

    1993-12-31

    Very few examples of non-destructive tests are available concerning of spent nuclear fuel elements after long period of dry storage under water. In Italy ENEL and FIAT CIEI performed two test campaigns in 1990 and 1991 at the pool storage facility AVOGADRO of Saluggia, aimed to investigate the condition of leaker B.W.R. fuel elements, dry-sealed into storage containers and stored under water since 1984. Radiochemical analyses were conducted on samples of the container`s cover gases by means of ``PSEUDO-SIPPING`` methods, with the following objectives: measurements of percentage of moisture radiolysis born hydrogen, detection of the possible presence of explosive mixtures; measurements of Kr 85 activity, verification of the behavior of cladding leaks. Results confirmed either the absence of dangerous quantities of radiolysis hydrogen, or a general increase of Kr 85 activity, compared with data coming from checks performed at the reactor site before fuel insertion into the storage containers. Cladding leaks at first were probably increased by transport conditions of spent fuel, dry-placed into shipping casks, and later on they were stabilized by the immersion in the pool cold water.

  7. 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.

  8. Wind-Aided Firespread Across Arrays of Discrete Fuel Elements

    DTIC Science & Technology

    1990-10-01

    Ph.D. thesis, Department of Chemical Engineering. Fredericton , Canada: University of New Brunswick. Fang, J. B., and Steward, F. R. 1969 Flame spread... Fredericton , Canada: University of New Brunswick. Steward, F. R., and Tennankore, K. N. 1981 The measurement of the burning rate of an individual dowel in a...1973 Flame spread through uniform fuel matrices. Report, Fire Science Center. Fredericton , Canada: University of New Brunswick. Steward, F. R

  9. 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.

  10. 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.

  11. 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.

  12. 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.

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

    SciTech Connect

    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.

  14. 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.

  15. 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.

  16. Transient Fission Gas Behavior in Uranium Nitride Fuel Under Proposed Space Applications

    DTIC Science & Technology

    1991-12-01

    functions of the detailed structure of the fuel pins and thus should provide reasonably accurate input to a detailed fuel modeling code. Calculations using...100 UN Fuel Grain Modelling Using Nonlinear Finite Element Analysis 101 By Capt. Daniel L. DeForest, USAF 102 103 Sample Input Deck 105 106...UN Fuel Grain Modelling Using Nonlinear Finite Element Analysis By Capt. Daniel L. DeForest, USAF Sample Input Deck CONDITIONS Power Density - 700.00

  17. 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.

  18. 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/.

  19. 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.

  20. 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].

  1. 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.

  2. Active neutron coincidence counting for the assay of MTR fuel elements

    SciTech Connect

    Sher, R.

    1983-02-01

    The active well coincidence counter (AWCC) and the neutron coincidence collar (CC) were investigated for their suitability to assay materials testing reactor (MTR) fuel elements. The AWCC was used with its special insert to hold the fuel element and interrogation source. The CC was modified by the addition of polyethylene liners 2.5 cm (1 in.) thick on the sides. For a typical MTR element (approx. 220 g /sup 235/U) and 1000-s count times, statistical errors were approx. 1.6% for the CC and approx. 0.6% for AWCC. For either instrument, the change in count rate corresponding to the removal or addition of one fuel plate (with an 18-plate element) was approx. 3.8%; thus, either instrument can detect removal of one plate. The AWCC can also detect removal of one plate in count times that are considerably less than 1000 s. Various functions were investigated to fit the coincidence count rate vs /sup 235/U mass curve for the AWCC. Programs have been written for the Hewlett-Packard HP-97 calculator to calculate the calibration constants of these functions by a least-squares technique. Coincidence count rates in the AWCC depend on the orientation of the plates of the fuel elements because of the counting efficiency variation in the insert. To lessen this dependence, the MTR element should be counted with its plates positioned vertically, that is, parallel to the radius of the device. For the collar, the effect of plate orientation is much smaller.

  3. 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.

  4. Preliminary study of nuclear fuel element testing based on coded source neutron imaging

    SciTech Connect

    Sheng Wang; Hang Li; Chao Cao; Yang Wu; Heyong Huo; Bin Tang

    2015-07-01

    Neutron radiography (NR) is one of the most important nondestructive testing methods, which is sensitive to low density materials. Especially, Neutron transfer imaging method could be used to test radioactivity materials refraining from γ effect, but it is difficult to realize tomography. Coded source neutron imaging (CSNI) is a newly NR method developed fast in the last several years. The distance between object and detector is much longer than traditional NR, which could be used to test radioactivity materials. With pre-reconstruction process from fold-cover projections, CSNI could easily realize tomography. This thesis carries out preliminary study on the nuclear fuel element testing by coded source neutron imaging. We calculate different enrichment, flaws and activity in nuclear fuel elements tested by CSNI with Monte-Carlo simulation. The results show that CSNI could be a useful testing method for nuclear fuel element testing. (authors)

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

    DOEpatents

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

    1957-09-10

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

  6. 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.

  7. 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.

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

    DOEpatents

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

    2011-03-01

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

  9. 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.

  10. 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.

  11. 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.

  12. Gene vector and transposable element behavior in mosquitoes.

    PubMed

    O'Brochta, David A; Sethuraman, Nagaraja; Wilson, Raymond; Hice, Robert H; Pinkerton, Alexandra C; Levesque, Cynthia S; Bideshi, Dennis K; Jasinskiene, Nijole; Coates, Craig J; James, Anthony A; Lehane, Michael J; Atkinson, Peter W

    2003-11-01

    The development of efficient germ-line transformation technologies for mosquitoes has increased the ability of entomologists to find, isolate and analyze genes. The utility of the currently available systems will be determined by a number of factors including the behavior of the gene vectors during the initial integration event and their behavior after chromosomal integration. Post-integration behavior will determine whether the transposable elements being employed currently as primary gene vectors will be useful as gene-tagging and enhancer-trapping agents. The post-integration behavior of existing insect vectors has not been extensively examined. Mos1 is useful as a primary germ-line transformation vector in insects but is inefficiently remobilized in Drosophila melanogaster and Aedes aegypti. Hermes transforms D. melanogaster efficiently and can be remobilized in this species. This element is also useful for creating transgenic A. aegypti, but its mode of integration in mosquitoes results in the insertion of flanking plasmid DNA. Hermes can be remobilized in the soma of A. aegypti and transposes using a common cut-and-paste mechanism; however, the element does not remobilize in the germ line. piggyBac can be used to create transgenic mosquitoes and occasionally integrates using a mechanism other than a simple cut-and-paste mechanism. Preliminary data suggest that remobilization is infrequent. Minos also functions in mosquitoes and, like the other gene vectors, appears to remobilize inefficiently following integration. These results have implications for future gene vector development efforts and applications.

  13. 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.

  14. 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.

  15. Microstructurally based finite element simulation of solder joint behavior

    SciTech Connect

    Frear, D.R.; Burchett, S.N.; Neilsen, M.K.; Stephens, J.J.

    1996-01-01

    The most commonly used solder for electrical interconnects in electronic packages is the near eutectic 60Sn-40Pb alloy. This alloy has a number of processing advantages (suitable melting point of 183C and good wetting behavior). However, under conditions of cyclic strain and temperature (thermomechanical fatigue) the microstructure of this alloy undergoes a heterogeneous coarsening and failure process that makes the prediction of solder joint lifetime complex. A finite element simulation methodology to predict solder joint mechanical behavior, that includes microstructural evolution, has been developed. The mechanical constitutive behavior was incorporated into the time dependent internal state variable viscoplastic model through experimental creep tests. The microstructural evolution is incorporated through a series of mathematical relations that describe mass flow in a temperature/strain environment. The model has been found to simulate observed thermomechanical fatigue behavior in solder joints.

  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. 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.

  18. 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.

  19. 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

  20. 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.

  1. Creep Behavior of Glass/Ceramic Sealant Used in Solid Oxide Fuel Cells

    SciTech Connect

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

    2010-01-02

    High operating temperature of solid oxide fuel cells require that sealant must function at high temperature between 600o and 900oC and in the oxidizing and reducing environments of fuel and air. It should be noted that creep deformation becomes relevant for a material when the operating temperature is near or exceeds half of its melting temperature (in degrees of Kelvin). The operating temperatures for most of the solid oxide fuel cells (SOFC) under development in the SECA program are around 800oC, which exceeds the glass transition temperature Tg for most glass ceramic materials. The goal of the study is to develop a creep model to capture the creep behavior of glass ceramic materials at high temperature and to investigate the effect of creep of glass ceramic sealant materials on stresses in glass seal and on the various interfaces of glass seal with other layers. The self-consistent creep models were incorporated into SOFC-MP and Mentat FC, and finite element analyses were performed to quantify the stresses in various parts. The stress in glass seals were released due to its creep behavior during the operating environments.

  2. 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.

  3. 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.

  4. 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.

  5. 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

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

    PubMed

    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.

  7. 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

  8. 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.

  9. 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

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

    SciTech Connect

    Gydesen, S.P.

    1993-07-01

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

  11. Aerothermal modeling program, Phase 2, Element C: Fuel injector-air swirl characterization

    NASA Technical Reports Server (NTRS)

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

    1987-01-01

    The main objectives of the NASA sponsored Aerothermal Modeling Program, Phase 2, Element C, are to collect benchmark quality data to quantify the fuel spray interaction with the turbulent swirling flows and to validate current and advanced two phase flow models. The technical tasks involved in this effort are discussed.

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

    NASA Technical Reports Server (NTRS)

    Wallace, Terry C.

    1991-01-01

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

  13. Elemental sulfur-producing high-temperature fuel gas desulfurization process

    SciTech Connect

    Anderson, G.L.; Garrigan, P.C.; Berry, F.O.

    1980-01-01

    Preliminary studies have shown that certain materials when added to air-regenerable, high-temperature, fuel gas desulfurization sorbents, such as iron oxide or zinc oxide, significantly increase elemental sulfur formation during regeneration. Although the full range of conditions under which these materials can be applied remains to be determined, successful applications could eliminate a costly SO/sub 2/ reduction step.

  14. 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.

  15. Behavior of Rare Earth Elements in Fractured Aquifers

    NASA Astrophysics Data System (ADS)

    Lee, S.; Kim, Y.; Lee, K.

    2003-12-01

    An understanding of the geochemistry of potential host rocks is very important in the site evaluation for construction of an underground geologic repository for radioactive waste. Because of similar valence and ionic radii and high similarity in electronic structure with trivalent actinides (such as Am3+ and Cm3+), the rare earth elements (REEs) have been used to predict the behavior of actinide-series elements in solution. For Am and Cm, which occur only in the trivalent states in most waste-disposal repository environments, the analogy with the REEs is particularly relevant. Krauskopf calculated the retardation factors for radionuclides in various rock materials based on some compiled data. But, in general, because the transuranic actinides do not occur naturally in appreciable quantities, their behaviors in repository environments cannot be predicted from evidence of their movement in geologic environments (mainly in groundwater) over geologic timespans. Predictions about long-term future behavior of transuranic actinides have therefore been made by extrapolation from short-term observations of their chemical properties in laboratory experiments or in field tests, but such extrapolation is fraught with uncertainty. In order to verify the behavior of Eu in various geological environments, we estimated the abundance of rare earth elements in three gneiss bodies originated from different geological environments and volcanic tuff. We also carried out some leaching experiment of fracture-filling calcite precipitated due to changes of geochemical environment in paleo-groundwater. Of the three gneisses, two gneisses are granitic-granodioritic origin and the other is tonaltic-trondjemitic origin. As a result, we could observe that Eu had a close relationship with fracture-filling calcite precipitation due to water-rock interaction. Our results show that Eu is the most variable element of REEs for the hydrogeological environment such as change of oxidation-reduction and

  16. 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.

  17. 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.

  18. 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.

  19. Comparison of the effect of insulating blockages on metal and oxide fuel elements

    SciTech Connect

    Tilbrook, R.W.; Dever, D.J.

    1988-01-01

    The safety philosophy of the new liquid-metal reactor (LMR) plant designs is oriented toward inherent protection against loss of coolable geometry and other entries to core disruption. One potential entry is via propagation of local faults. The basic event in all local sequences is cladding failure, irrespective of initiator. A model of a complete insulating blockage, i.e., total loss of heat transfer from the cladding surface due to any cause, was developed for a range of insulated arcs. The internal properties represented either metal or oxide fuels, both irradiated to a condition that closed the fuel-clad gap. The advantage of the high conductivity of the metal fuel is clearly evident; the maximum cladding temperatures are considerably lower than for the oxide elements with the same circumferential blockage extent. Also, the minimum cladding temperature at the opposite side of the element is higher for the metal fuel, thus providing more uniform heat rejection from the unblocked portion of the cladding. The cladding temperatures at the edge of the blockages for the oxide elements are directly proportional to the blockage angle, indicating that the cladding is the main path for heat rejection.

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

    NASA Astrophysics Data System (ADS)

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

    1989-07-01

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

  1. 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

  2. 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.

  3. 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.

  4. 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.

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

    PubMed

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

    2015-10-30

    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.

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

    NASA Technical Reports Server (NTRS)

    Schneider, Steven J. (Inventor)

    2002-01-01

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

  7. 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.

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

    NASA Technical Reports Server (NTRS)

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

    1986-01-01

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

  9. 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

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

    SciTech Connect

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

    1982-05-01

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

  11. 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.

  12. 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.

  13. 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.

  14. Modeling bistable behaviors in morphing structures through finite element simulations.

    PubMed

    Guo, Qiaohang; Zheng, Huang; Chen, Wenzhe; Chen, Zi

    2014-01-01

    Bistable structures, exemplified by the Venus flytrap and slap bracelets, can transit between different configurations upon certain external stimulation. Here we study, through three-dimensional finite element simulations, the bistable behaviors in elastic plates in the absence of terminate loads, but with pre-strains in one (or both) of the two composite layers. Both the scenarios with and without a given geometric mis-orientation angle are investigated, the results of which are consistent with recent theoretical and experimental studies. This work can open ample venues for programmable designs of plant/shell structures with large deformations, with applications in designing bio-inspired robotics for biomedical research and morphing/deployable structures in aerospace engineering.

  15. Studies on disintegrating spherical fuel elements of high temperature gas-cooled reactor by a electrochemical method

    NASA Astrophysics Data System (ADS)

    Tian, Lifang; Wen, Mingfen; Chen, Jing

    2013-01-01

    Spherical fuel elements of a high temperature gas-cooled reactor were disintegrated through a electrochemical method with NaNO3 as electrolyte. The X-ray diffraction spectra and total carbon contents of the graphite fragments were determined, and the results agreed with those from simulated fuel elements. After conducting the characterization analysis and the leaching experiment of coated fuel particles, the uranium concentrations of leaching solutions and spent electrolyte were found to be at background levels. The results demonstrate the effectiveness of the improved electrochemical method with NaNO3 as electrolyte in disintegrating the unirradiated fuel elements without any damage to the coated fuel particles. Moreover, the method avoided unexpected radioactivity contamination to the graphite matrix and spent electrolyte.

  16. Behavior of Zr1%Nb Fuel Cladding under Accident Conditions

    SciTech Connect

    Perez-Fero, E.; Hozer, Z.; Windberg, P.; Nagy, I.; Vimi, A.; Ver, N.; Matus, L.; Kunstar, M.; Novotny, T.; Horvath, M.; Gyori, Cs.

    2007-07-01

    The behavior of the VVER fuel (E110) cladding under accident conditions has been investigated at the AEKI in order to study the role of oxidation and hydrogen uptake on the cladding embrittlement and to understand the phenomena that took place during the Paks-2 cleaning tank incident (2003). The test programme covered small scale tests and large scale tests with electrically heated 7-rod bundles in the CODEX (Core Degradation Experiment) facility. Since a hydrogen rich atmosphere could have been formed in the closed tank, the experiments were carried out in hydrogen-steam mixture. According to the results of the small scale tests, a former correlation for the ductile-brittle transitions of E110 in pure steam remained valid in hydrogen rich steam atmosphere as well. During the large scale tests the main conditions of the incident were reconstructed. The test characterized the high temperature oxidation and embrittlement of zirconium in hydrogen rich steam. The observed cladding failure phenomena and the extent of the damage of the test bundle in the quenching phase were very similar to those of the VVER assemblies in the incident. The simulation of the cleaning tank incident provided detailed information on the most probable scenario of the incident. (authors)

  17. Reducing uncertainties affecting the assessment of the long-term corrosion behavior of spent nuclear fuel.

    PubMed

    Fanghänel, Thomas; Rondinella, Vincenzo V; Glatz, Jean-Paul; Wiss, Thierry; Wegen, Detlef H; Gouder, Thomas; Carbol, Paul; Serrano-Purroy, Daniel; Papaioannou, Dimitrios

    2013-04-01

    Reducing the uncertainties associated with extrapolation to very long term of corrosion data obtainable from laboratory tests on a relatively young spent nuclear fuel is a formidable challenge. In a geologic repository, spent nuclear fuel may come in contact with water tens or hundreds of thousands of years after repository closure. The corrosion behavior will depend on the fuel properties and on the conditions characterizing the near field surrounding the spent fuel at the time of water contact. This paper summarizes the main conclusions drawn from multiyear experimental campaigns performed at JRC-ITU to study corrosion behavior and radionuclide release from spent light water reactor fuel. The radionuclide release from the central region of a fuel pellet is higher than that from the radial periphery, in spite of the higher burnup and the corresponding structural modifications occurring at the pellet rim during irradiation. Studies on the extent and time boundaries of the radiolytic enhancement of the spent fuel corrosion rate indicate that after tens or hundreds of thousands of years have elapsed, very small or no contribution to the enhanced corrosion rate has to be expected from α radiolysis. A beneficial effect inhibiting spent fuel corrosion due to the hydrogen overpressure generated in the near field by iron corrosion is confirmed. The results obtained so far point toward a benign picture describing spent fuel corrosion in a deep geologic repository. More work is ongoing to further reduce uncertainties and to obtain a full description of the expected corrosion behavior of spent fuel.

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

    SciTech Connect

    Sarta, Jose A.; Castiblanco, Luis A

    2005-05-24

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

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

    NASA Astrophysics Data System (ADS)

    Sarta, Josè A.; Castiblanco, Luis A.

    2005-05-01

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

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

    Code of Federal Regulations, 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..., grinding and grading will be present. Mixed oxide fuels are handled in glove boxes (or...

  1. Test plan for techniques to measure and remove coatings from K West Basin fuel elements

    SciTech Connect

    Bridges, A.E.

    1998-06-17

    Several types of coatings have previously been visually identified on the surface of 105-K East and 105-K West Basins fuel elements. One type of coating (found only in K West Basin) in particular was found to be a thick translucent material that was often seen to be dislodged from the elements as flakes when the elements were handled during visual examinations (Pitner 1997). Subsequently it was determined (for one element only in a hot cell) that this material, in the dry condition, could easily be removed from the element using a scraping tool. The coating was identified as Al(OH){sub 3} through X-ray diffraction (XRD) analyses and to be approximately 60 {micro}m thick via scanning electron microscopy (SEM). However, brushing under water in the basin using numerous mechanical strokes failed to satisfactorily remove these coatings in their thickest form as judged by appearance. Such brushing was done with only one type of metal brush, a brush design previously found satisfactory for removing UO{sub 4}.xH{sub 2}O coatings from the elements.

  2. 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

  3. 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

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

    NASA Astrophysics Data System (ADS)

    Sarta Fuentes, Jose Antonio

    2005-04-01

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

  5. 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

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

    SciTech Connect

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

    1999-07-01

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

  7. Consumer behavior towards fuel efficient vehicles. Volume I: executive summary. Final report Oct 77-Feb 80

    SciTech Connect

    Sherman, L.; Manski, C.F.; Ginn, J.R.

    1980-02-01

    This report assesses consumer behavior towards fuel-efficient vehicles designed to meet recently mandated federal fuel economy standards. The study involves a comprehensive evaluation of existing nationwide survey data as well as the development of a major new econometric forecasting model of household vehicle type choice. As a result, the report describes both an assessment of consumers' current reported sentiments toward fuel-efficient vehicles and insights into expected future changes in household vehicle purchase behavior in response to changes in vehicle designs and prices, demographics, and the energy environment.

  8. Modeling and simulation of the dynamic behavior of a polymer electrolyte membrane fuel cell

    NASA Astrophysics Data System (ADS)

    Yerramalla, Sampath; Davari, Asad; Feliachi, Ali; Biswas, Tamal

    The focus of this paper is to develop a mathematical model for investigating the dynamic performance of a polymer electrolyte membrane fuel cell. The model in this work is based on physical laws having clear significance in replicating the fuel cell system and can easily be used to set up different operational strategies. Simulation results display the transient behavior of the voltage within each single cell, and also within a number of such single cells combined into a fuel cell stack system. A linear as well as a nonlinear analysis of the polymer electrolyte membrane fuel cell system has been discussed in order to present a complete and comprehensive view of this kind of modeling. Also, a comparison of the two kinds of analysis has been performed. Finally, the various characteristics of the fuel cell system are plotted in order to help us understand its dynamic behavior. Results indicate that there is a considerable amount of error in the modeling process if we use a linear model of the fuel cell. Thus, the nonlinearities present in the fuel cell system should be taken into account in order to obtain a better understanding of the dynamic behavior of the fuel cell system.

  9. 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.

  10. Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

    SciTech Connect

    R. L. Williamson

    2011-08-01

    A powerful multidimensional fuels performance analysis 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. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

  11. Double-Fueled Janus Swimmers with Magnetotactic Behavior.

    PubMed

    Schattling, Philipp S; Ramos-Docampo, Miguel A; Salgueiriño, Verónica; Städler, Brigitte

    2017-03-29

    Self-propelled particles attract a great deal of attention due to the auspicious range of applications for which nanobots can be used. In a biomedical context, self-propelled swimmers hold promise to autonomously navigate to a desired location in an attempt to counteract cell/tissue defects either by releasing drugs or by performing surgical tasks. The vast majority of prior reports deal with single engine assemblies, often utilizing fuel molecules which are considered to be highly cytotoxic. Herein, we introduce two engines: (1) a motor which couples enzymes (i.e., glucose oxidase) and inorganic nanoparticles (i.e., platinum nanoparticles) to gain power and (2) a peptide-fueled trypsin motor. We demonstrate that both engines can induce enhanced diffusion properties of (Janus) particles using bioavailable and completely harmless fuel molecules. By combining both engines on the same carrier, we show self-propelled particles employing two independent engines, using two different fuels. A collaborative enhancement of the swimmer's diffusion properties upon powering-up both engines simultaneously is observed. Additionally, the incorporation of magnetic nanoparticles allows for the swimmer to move in a magnetic gradient upon applying an external magnetic field, yielding in directional motion of the double-fueled particles. These multiple-fueled biocompatible swimmers are a significant contribution to make them applicable in a biomedical context.

  12. 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.

  13. 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.

  14. 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.

  15. Use of molybdenum as a structural material of fuel elements for improving the safety of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Shmelev, A. N.; Kozhahmet, B. K.

    2017-01-01

    Main purpose of the study is justifying the use of molybdenum as a structural material of fuel elements for improving the safety of nuclear reactors. Particularity of used molybdenum is that its isotopic composition corresponds to molybdenum, which is obtained as the tailing during operation of the separation cascade for producing a material for medical diagnostics of cancer. When performing the study the neutron-physical properties of isotopes of natural molybdenum (nuclear data library JENDL-4.0) and thermal properties of metallic molybdenum were used. The following results were obtained: 1. A method for reducing the thermal constant of fuel elements for light water and fast reactors by using dispersion fuel in cylindrical fuel rods containing, for example, granules of metallic U-Mo-alloy into Mo-matrix was proposed. 2. The necessity of molybdenum enrichment by weakly absorbing isotopes was shown. 3. Total use of isotopic molybdenum will be more than 50%. A method for reducing the thermal constant of the fuel elements, allowing us to increase the safety of light water and fast nuclear reactors by using dispersion fuel in cylindrical fuel rods containing, for example, granules of metallic U-Mo-alloy into Mo-matrix with enrichment by weakly absorbing isotopes of molybdenum is proposed.

  16. 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

  17. 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

  18. Model development of a polymer electrolyte membrane fuel cell to predict steady and unsteady behavior

    NASA Astrophysics Data System (ADS)

    Mishra, Bikash

    Fuel cells are promising technology to meet the energy need of the future. This alternative energy source is clean and efficient, and with the continuous decrease in fossil fuel resources, one of the best bets towards sustaining our power needs. Fuel cells are being used in automobiles as well as to fulfill portable power needs. In this work a computational model has been developed for fuel cells which can be used to simulate traditional as well as passive proton exchange membrane fuel cell behavior. The model is unsteady, two phase, nonisothermal in nature, and also capable of handling natural convection or buoyancy driven flows. The model also takes into account electrochemical reactions at catalyst sites. The model has been implemented and validated against experiments. It is used to carry out unsteady simulations to study start-up characteristic of proton exchange membrane fuel cells and to follow the behavior of liquid water as well as heat transfer within the cell. The buoyancy model is used to simulate a natural convection region and a passive fuel cell (used for portable applications). Design of passive fuel cells is driven by high temperature regimes and that issue has been further explored.

  19. 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.

  20. Spent fuel reaction - the behavior of the {epsilon}-phase over 3.1 years

    SciTech Connect

    Finn, P.A.; Hoh, J.C.; Wolf, S.F.

    1996-12-31

    The release fractions of the five elements in the {epsilon}-phase ({sup 99}Tc, {sup 97}Mo, Ru, Rh, and Pd) as well as that of {sup 238}U are reported for the reaction of two oxide fuels (ATM-103 and ATM-106) in unsaturated tests under oxidizing conditions. The {sup 99}Tc release fractions provide a lower limit for the magnitude of the spent fuel reaction. The {sup 99}Tc release fractions indicate that a surface reaction might be the rate controlling mechanism for fuel reaction under unsaturated conditions and the oxidant is possibly H{sub 2}O{sub 2}, a product of alpha radiolysis of water.

  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. Outline for a multi-cell nuclear thermionic fuel element that may be pretested with electric heat

    NASA Astrophysics Data System (ADS)

    Wilson, Volney C.

    1997-01-01

    A nuclear thermionic converter electrical generating system is proposed in which the nuclear fuel is clad in tungsten (W) and transmits heat to a tungsten emitter by radiation. The tungsten clad is a single unit, containing a continuous fuel stack with an unfueled section extending through one end of the reactor. The emitters are electrically insulated from the heat source; therefore, several converters may be connected by short leads to produce more voltage per fuel element and to reduce the power losses in the leads. A fast reactor design was chosen; consequently, tungsten may be used for the fuel cladding and the emitters without a significant reactivity penalty due to neutron capture by tungsten epithermal resonances. The ability to use all-tungsten emitters may permit high emitter temperatures. Calculations indicate that at an emitter temperature of 2150 K and current density of 10 A/cm2, a 36 cm long thermionic fuel element (TFE) with 9 converters in series should produce 4500 We at 9.2 V and 15.7% efficiency. One major advantage of this approach, relative to typical multicell designs is that the system can be tested by electrical heaters in the fuel cavity before loading fuel.

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

    DOEpatents

    Mao, Chien-Pei [Clive, IA; Short, John [Norwalk, IA; Klemm, Jim [Des Moines, IA; Abbott, Royce [Des Moines, IA; Overman, Nick [West Des Moines, IA; Pack, Spencer [Urbandale, IA; Winebrenner, Audra [Des Moines, IA

    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.

  4. 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.}

  5. Experimental Investigation of Vibratory Stresses in a Concentric-Ring Direct-Air-Cycle Nuclear Fuel Element

    NASA Technical Reports Server (NTRS)

    Chiarito, Patrick T.

    1957-01-01

    Preliminary tests made by the General Electric Company indicated that aerodynamic loads might cause large enough distortions in the thin sheet-metal rings of a nuclear fuel element to result in structural failure. The magnitude of the distortions in a test fuel element was determined from strains measured with airflow conditions simulating those expected during engine operation. The measured vibratory strains were low enough to indicate the improbability of failure by fatigue. A conservative estimate of the radial deflection that accompanied peak strains in the outer ring was +0.0006 inch.

  6. The Vaporization Behavior of a Fuel Drop on a Hot Surface

    DTIC Science & Technology

    1977-11-01

    evaporation behavior of fuel drops 99 Figure 36. Effect of surface cleanliness on drop evaporation lifetime .. ......... . 101 Figure 37. Effect of drop...C, CD 0 C) - -C) -H _______________C_ 100 procedures that were considered during the evaluation included the surface cleanliness , fuel drop size and...evaporating surface heating rate. The effect of the studied variables on the test results was found to be as follow: Surface Cleanliness As indicated

  7. Effects of fuel particle size and fission-fragment-enhanced irradiation creep on the in-pile behavior in CERCER composite pellets

    NASA Astrophysics Data System (ADS)

    Zhao, Yunmei; Ding, Shurong; Zhang, Xunchao; Wang, Canglong; Yang, Lei

    2016-12-01

    The micro-scale finite element models for CERCER pellets with different-sized fuel particles are developed. With consideration of a grain-scale mechanistic irradiation swelling model in the fuel particles and the irradiation creep in the matrix, numerical simulations are performed to explore the effects of the particle size and the fission-fragment-enhanced irradiation creep on the thermo-mechanical behavior of CERCER pellets. The enhanced irradiation creep effect is applied in the 10 μm-thick fission fragment damage matrix layer surrounding the fuel particles. The obtained results indicate that (1) lower maximum temperature occurs in the cases with smaller-sized particles, and the effects of particle size on the mechanical behavior in pellets are intricate; (2) the first principal stress and radial axial stress remain compressive in the fission fragment damage layer at higher burnup, thus the mechanism of radial cracking found in the experiment can be better explained.

  8. 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.

  9. 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

  10. 3D laser inspection of fuel assembly grid spacers for nuclear reactors based on diffractive optical elements

    NASA Astrophysics Data System (ADS)

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

    2007-06-01

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

  11. 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.

  12. 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.}

  13. 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.

  14. 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

  15. 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

  16. 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.

  17. Thermal-hydraulic analysis of an annular fuel element: The Achilles' heel of the particle bed reactor

    SciTech Connect

    Dibben, M.J.; Tuttle, R.F. )

    1993-01-20

    The low pressure nuclear thermal propulsion (LPNTP) concept offers significant improvements in rocket engine specific impulse over rockets employment chemical propulsion. This study investigated a parametric thermal-hydraulic analysis of an annular fueld element, also referred to as a fuel pipe, using the computer code ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer). The fuelpipe is an annular particle bed fuel element of the reactor with radially inward flow of hydrogen through the element. In this study, the outlet temperature of the hydrogen is parametrically related to key effects, including the reactor power at two different pressure drops, the effect of power coupling for in-core testing, and the effect of hydrogen flow rates. Results show that the temperature is linearly related to the reactor power, but not to pressure drop, and that cross flow inside the fuelpipe occurs at approximately 0.3 percent of the radial flow rates.

  18. 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)

  19. 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.

  20. Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

    SciTech Connect

    Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng; Alankar, Alankar; Subramanian, Gopinath; Stanek, Christopher

    2017-01-01

    Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. CASL has endeavored to improve upon this approach by incorporating a microstructurally-based, atomistically-informed, zirconium alloy mechanical deformation analysis capability into the BISON-CASL engineering scale fuel performance code. Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tome´ [2], has been coupled with BISON-CASL to represent the mechanistic material processes controlling the deformation behavior of the cladding. A critical component of VPSC is the representation of the crystallographic orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON-CASL and provides initial results utilizing the coupled functionality.

  1. Common Elements for the Psychotherapeutic Management of Patients with Self Injurious Behavior

    ERIC Educational Resources Information Center

    Schiavone, Francesca L.; Links, Paul S.

    2013-01-01

    Objectives: Current research suggests that effective psychotherapies for Self Injurious Behavior (SIB) in the context of Borderline Personality Disorder (BPD) contain generic common elements which are responsible for their success. Because of the links between BPD, SIB, and child abuse, it is likely that these common elements can also be applied…

  2. 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.

  3. 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

  4. 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.

  5. 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.

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

    SciTech Connect

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

    1980-01-01

    Variable width ribbon heating elements have been fabricated which provide a chopped cosine, variable heat flux profile for fuel rod 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 result in the derivation of an analytical expression for the ribbon contours. From this, the ribbons are machined and wound on numerically controlled equipment. Postprocessing and inspection results in a wound, variable width ribbon with the precise dimensional, electrical, and mechanical properties needed for use in fuel pin simulators.

  7. 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.

  8. 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.

  9. Behavior of Radionuclides and RCRA Elements in Tank Backfill Grouts

    SciTech Connect

    Fuhrmann, M.; Heiser, J.; Gillow, J.

    2006-07-01

    One approach to decommissioning emptied high-level waste tanks is to backfill them with grout. Because of the long lives and high toxicity of some of the contaminants, the chemical behavior of the contaminants in the grout need to be understood, especially as the grout ages and weathers over long times. In this paper, the sequestration of technetium and iodine in contact with two grout formulations, and their component materials, is discussed. Preliminary results are presented of experiments examining the solubility of actinides in contact with the grouts as pH is lowered and carbonate content increased, representing conditions of a weathered grout system. (authors)

  10. Modeling the behavior of optical elements in radiation environments

    SciTech Connect

    Barlow, T.A.; Rhoades, C.E. Jr.; Merker, M.; Triplett, J.R.

    1986-11-03

    Calculation of heating caused by the deposition of x-rays in thin film optical elements is complicated because the mean free path of photo and autoionization electrons is comparable to the thin film thickness and thus the electron deposition cannot be considered local. This paper describes the modeling in a 1-D code of: (a) x-ray deposition and transport; (b) electron production, deposition and transport; and (c) thermal conduction and transport. X-ray transport is handled by multigroup discrete ordinates, electron transport is done by the method of characteristics, applied to the two term spherical harmonics expansion approximation (P1) to the Spencer-Lewis transport equation, and thermal transport is computed by a simple Richardson extrapolation of a backward Euler solution to the heat conduction equations. Results of a few test cases are presented. 8 refs., 26 figs., 2 tabs.

  11. 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.

  12. 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

  13. 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.

  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. Comparison of particle size distributions and elemental partitioning from the combustion of pulverized coal and residual fuel oil.

    PubMed

    Linak, W P; Miller, C A; Wendt, J O

    2000-08-01

    U.S. Environmental Protection Agency (EPA) research examining the characteristics of primary PM generated by the combustion of fossil fuels is being conducted in efforts to help determine mechanisms controlling associated adverse health effects. Transition metals are of particular interest, due to the results of studies that have shown cardiopulmonary damage associated with exposure to these elements and their presence in coal and residual fuel oils. Further, elemental speciation may influence this toxicity, as some species are significantly more water-soluble, and potentially more bio-available, than others. This paper presents results of experimental efforts in which three coals and a residual fuel oil were combusted in three different systems simulating process and utility boilers. Particle size distributions (PSDs) were determined using atmospheric and low-pressure impaction as well as electrical mobility, time-of-flight, and light-scattering techniques. Size-classified PM samples from this study are also being utilized by colleagues for animal instillation experiments. Experimental results on the mass and compositions of particles between 0.03 and > 20 microns in aerodynamic diameter show that PM from the combustion of these fuels produces distinctive bimodal and trimodal PSDs, with a fine mode dominated by vaporization, nucleation, and growth processes. Depending on the fuel and combustion equipment, the coarse mode is composed primarily of unburned carbon char and associated inherent trace elements (fuel oil) and fragments of inorganic (largely calcium-alumino-silicate) fly ash including trace elements (coal). The three coals also produced a central mode between 0.8- and 2.0-micron aerodynamic diameter. However, the origins of these particles are less clear because vapor-to-particle growth processes are unlikely to produce particles this large. Possible mechanisms include the liberation of micron-scale mineral inclusions during char fragmentation and burnout

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

    NASA Astrophysics Data System (ADS)

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

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

  17. 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.

  18. Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

    NASA Astrophysics Data System (ADS)

    Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng; Alankar, Alankar; Subramanian, Gopinath; Stanek, Christopher

    2017-01-01

    Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

  19. Meta-analysis of treatment for child sexual behavior problems: practice elements and outcomes.

    PubMed

    St Amand, Annick; Bard, David E; Silovsky, Jane F

    2008-05-01

    This meta-analysis of 11 treatment outcome studies evaluated 18 specific treatments of sexual behavior problems (SBP) as a primary or secondary target. Specifically, it examines relations among child characteristics, treatment characteristics (including practice elements), and short-term outcome (including sexual and general behavior problems). Utilizing pre- and postintervention results, the overall degree of change over the course of treatment was estimated at a 0.46 and 0.49 standard deviation decline in SBP and general behavior problems, respectively. As hypothesized, the caregiver practice element Parenting/Behavior Management Skills (BPT) predicted the Child Sexual Behavior Inventory (and the Child Behavior Checklist when BPT was combined with caregiver Rules about Sexual Behaviors). In contrast, practice elements that evolved from Adult Sex Offender (ASO) treatments were not significant predictors. BPT and preschool age group provided the best model fit and more strongly predicted outcome than broad treatment type classifications (e.g., Play Therapy or Cognitive Behavior Therapy). Results question current treatments for children with SBP that are based on ASO models of treatment without caregiver involvement.

  20. Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel

    NASA Astrophysics Data System (ADS)

    Jeong, G. Y.; Kim, Yeon Soo; Jamison, L. M.; Robinson, A. B.; Lee, K. H.; Sohn, Dong-Seong

    2017-04-01

    U-Mo/Al dispersion fuel irradiated to high burnup at high power (high fission rate) exhibited microstructural changes including deformation of the fuel particles, pore growth, and rupture of the Al matrix. The driving force for these microstructural changes was meat swelling resulting from a combination of fuel particle swelling and interaction layer (IL) growth. In some cases, pore growth in the interaction layers also contributed to meat swelling. The main objective of this work was to determine the stress distribution within the fuel meat that caused these phenomena. A mechanical equilibrium between the stress generated by fuel meat swelling and the stress relieved by fission-induced creep in the meat constituents (U-Mo particles, Al matrix, and IL) was considered. Test plates with well-recorded fabrication data and irradiation conditions were used, and their post-irradiation examination (PIE) data was obtained. ABAQUS finite element analysis (FEA) was utilized to simulate the microstructural evolution of the plates. The simulation results allowed for the determination of effective stress and hydrostatic stress exerted on the meat constituents. The effects of fabrication and irradiation parameters on the stress distribution that drives microstructural evolutions, such as pore growth in the IL and Al matrix rupture, were investigated.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

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

    DOE PAGES

    Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; ...

    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

  7. Impacts of climate on shrubland fuels and fire behavior in the Owyhee Basin, Idaho

    NASA Astrophysics Data System (ADS)

    Vogelmann, J. E.; Shi, H.; Hawbaker, T.; Li, Z.

    2013-12-01

    There is evidence that wildland fire is increasing as a function of global change. However, fire activity is spatially, temporally and ecologically variable across the globe, and our understanding of fire risk and behavior in many ecosystems is limited. After a series of severe fire seasons that occurred during the late 1990's in the western United States, the LANDFIRE program was developed with the goals of providing the fire community with objective spatial fuel data for assessing wildland fire risk. Even with access to the data provided by LANDFIRE, assessing fire behavior in shrublands in sagebrush-dominated ecosystems of the western United States has proven especially problematic, in part due to the complex nature of the vegetation, the variable influence of understory vegetation including invasive species (e.g. cheatgrass), and prior fire history events. Climate is undoubtedly playing a major role, affecting the intra- and inter-annual variability in vegetation conditions, which in turn impacts fire behavior. In order to further our understanding of climate-vegetation-fire interactions in shrublands, we initiated a study in the Owyhee Basin, which is located in southwestern Idaho and adjacent Nevada. Our goals include: (1) assessing the relationship between climate and vegetation condition, (2) quantifying the range of temporal variability in grassland and shrubland fuel loads, (3) identifying methods to operationally map the variability in fuel loads, and (4) assessing how the variability in fuel loads affect fire spread simulations. To address these goals, we are using a wide variety of geospatial data, including remotely sensed time-series data sets derived from MODIS and Landsat, and climate data from DAYMET and PRISM. Remotely-sensed information is used to characterize climate-induced temporal variability in primary productivity in the Basin, where fire spread can be extensive after senescence when dry vegetation is added to dead fuel loads. Gridded

  8. 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.

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

    SciTech Connect

    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.

  10. 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.

  11. 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.

  12. Effects of Fuel Spray Modeling on Combustion Instability Predictions in a Single-Element Lean Direct Injection (LDI) Gas Turbine Combustor

    DTIC Science & Technology

    2014-09-01

    Modeling on Combustion Instability Predictions in a Single- Element Lean Direct Injection (LDI) Gas Turbine Combustor 5b. GRANT NUMBER 5c. PROGRAM...injection (LDI) gas turbine combustor were performed to assess the effects of fuel spray modeling in predicting combustion instability. Concurrently...239.18 Effects of Fuel Spray Modeling on Combustion Instability Predictions in a Single-Element Lean Direct Injection (LDI) Gas Turbine

  13. Simulation of the irradiation-induced micro-thermo-mechanical behaviors evolution in ADS nuclear fuel pellets

    NASA Astrophysics Data System (ADS)

    Ding, Shurong; Zhao, Yunmei; Wan, Jibo; Gong, Xin; Wang, Canglong; Yang, Lei; Huo, Yongzhong

    2013-11-01

    An Accelerator Driven System (ADS) is dedicated to Minor Actinides (MA) transmutation. The fuels for ADS are highly innovative, which are composite fuel pellets with the fuel particles containing MA phases dispersed in a MgO or Mo matrix. Assuming that the fuel particles are distributed periodically in the MgO matrix, a three-dimensional finite element model is developed. The three-dimensional incremental large-deformation constitutive relations for the fuel particles and matrix are separately built, and a method is accordingly constructed to implement simulation of the micro-thermo-mechanical behaviors evolution. Evolutions of the temperature and mechanical fields are given and discussed. With irradiation creep included in the MgO matrix constitutive relation, the conclusions can be drawn as that (1) irradiation creep has a remarkable effect on the mechanical behaviors evolution in the matrix; (2) irradiation creep plays an important role in the damage mechanism interpretation of ceramic matrix fuel pellets. Thermal conductivity The thermal conductivity model is adopted as KUO2 = K0·FD·FP·FM·FR, which was proposed by Lucuta et al. [10] to adapt to the high burnup conditions with consideration of the effects of temperature, burnup, porosity and fission products. K0 is the thermal conductivity of fully dense un-irradiated UO2, as Eq. (1) in W/m K; FD, FP are the adjust factors reflecting the effects of dissolved and precipitated fission products; FM and FR are factors due to porosity and irradiation effects. The adopted thermal conductivity varies with temperature and burnup, which expresses its degradation with burnup, with the terms as k0={1}/{0.0375+2.165×10-4T}+{4.715×109}/{T2}exp-{16361}/{T} FD={1.09}/{B3.265}+{0.0643}/{√{B}}√{T}artan{1}/{1.09/B3.265}+{0.0643}/{√{B}}√{T} FP=1+0.019B/3-0.019B{1}/{1+exp(1200-T100)} FM={1-P}/{1+(s-1)P} FR=1-{0.2}/{1+expT-90080} Thermal expansion The engineering strain of thermal expansion [11] is given as {ΔL}/{L0

  14. Identifying Common Elements of Evidence-Based Psychosocial Treatments for Children's Disruptive Behavior Problems

    ERIC Educational Resources Information Center

    Garland, Ann F.; Hawley, Kristin M.; Brookman-Frazee, Lauren; Hurlburt, Michael S.

    2008-01-01

    The article reviews a new method to identify common core elements of evidence-based practice (EBP) for children with disruptive behavior problems and their parents. This will not only help in characterizing practice, but also will enhance the skills and attitudes of therapists and be useful in implementing EBP in community-based service settings.

  15. Behavior or Nonmetallic Materials in Shale Oil Derived Jet Fuels and in High Aromatic and High Sulfur Petroleum Fuels

    DTIC Science & Technology

    1978-07-01

    Sealants (Noncuring Type) 30 6. Elastomeric Marmon Clamp Seals 34 7. Elastomeric O-Ring Materials 38 8. Fuel Cell Baffle Materials 42 9. Internal Fuel Cell...fuel bladder materials, bladder repair adhesives, groove sealants, elastomeric o-ring materials, elasto- meric Marmon clamp seals, fuel tank coatings...a. EC-2216 EC-2216 is a room temperature curing epoxy- polyamide struc- tural adhesive manufactured by 3M Company. The initial test results have

  16. 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.

  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. Feasibility Study for Monitoring Actinide Elements in Process Materials Using FO-LIBS at Advanced spent fuel Conditioning Process Facility

    SciTech Connect

    Han, Bo-Young; Choi, Daewoong; Park, Se Hwan; Kim, Ho-Dong; Dae, Dongsun; Whitehouse, Andrew I.

    2015-07-01

    Korea Atomic Energy Research Institute (KAERI) have been developing the design and deployment methodology of Laser- Induced Breakdown Spectroscopy (LIBS) instrument for safeguards application within the argon hot cell environment at Advanced spent fuel Conditioning Process Facility (ACPF), where ACPF is a facility being refurbished for the laboratory-scaled demonstration of advanced spent fuel conditioning process. LIBS is an analysis technology used to measure the emission spectra of excited elements in the local plasma of a target material induced by a laser. The spectra measured by LIBS are analyzed to verify the quality and quantity of the specific element in the target matrix. Recently LIBS has been recognized as a promising technology for safeguards purposes in terms of several advantages including a simple sample preparation and in-situ analysis capability. In particular, a feasibility study of LIBS to remotely monitor the nuclear material in a high radiation environment has been carried out for supporting the IAEA safeguards implementation. Fiber-Optic LIBS (FO-LIBS) deployment was proposed by Applied Photonics Ltd because the use of fiber optics had benefited applications of LIBS by delivering the laser energy to the target and by collecting the plasma light. The design of FO-LIBS instrument for the measurement of actinides in the spent fuel and high temperature molten salt at ACPF had been developed in cooperation with Applied Photonics Ltd. FO-LIBS has some advantages as followings: the detectable plasma light wavelength range is not limited by the optical properties of the thick lead-glass shield window and the potential risk of laser damage to the lead-glass shield window is not considered. The remote LIBS instrument had been installed at ACPF and then the feasibility study for monitoring actinide elements such as uranium, plutonium, and curium in process materials has been carried out. (authors)

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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.

  5. Experimental Combustion Dynamics Behavior of a Multi-Element Lean Direct Injection (LDI) Gas Turbine Combustor

    NASA Technical Reports Server (NTRS)

    Acosta, Waldo A.; Chang, Clarence T.

    2016-01-01

    An experimental investigation of the combustion dynamic characteristics of a research multi-element lean direct injection (LDI) combustor under simulated gas turbine conditions was conducted. The objective was to gain a better understanding of the physical phenomena inside a pressurized flametube combustion chamber under acoustically isolated conditions. A nine-point swirl venturi lean direct injection (SV-LDI) geometry was evaluated at inlet pressures up to 2,413 kPa and non-vitiated air temperatures up to 867 K. The equivalence ratio was varied to obtain adiabatic flame temperatures between 1388 K and 1905 K. Dynamic pressure measurements were taken upstream of the SV-LDI, in the combustion zone and downstream of the exit nozzle. The measurements showed that combustion dynamics were fairly small when the fuel was distributed uniformly and mostly due to fluid dynamics effects. Dynamic pressure fluctuations larger than 40 kPa at low frequencies were measured at 653 K inlet temperature and 1117 kPa inlet pressure when fuel was shifted and the pilot fuel injector equivalence ratio was increased to 0.72.

  6. Constitutive Behavior of Reinforced Concrete Membrane Elements under Tri-directional Shear

    NASA Astrophysics Data System (ADS)

    Labib, Moheb

    The two-dimensional behavior of typical reinforced concrete (RC) structures has been extensively studied in the past several decades by investigating the constitutive behavior of full-scale reinforced concrete elements subjected to a bi-axial state of stress. In order to understand the true behavior of many large complex structures, the goal of this investigation is to develop new constitutive relationships for RC elements subjected to tri-directional shear stresses. Recently, additional out-of-plane jacks were installed on the panel tester at University of Houston so that concrete elements could be subjected to tri-directional shear stresses. This upgrade makes the panel tester the only one of its kind in the US that is capable of applying such combinations of stresses on full-scale reinforced concrete elements. This dissertation presents the details of the mounting and installation of the additional hydraulic jacks on the universal panel tester. The experimental program includes a series of seven reinforced concrete elements subjected to different combinations of in-plane and out-of-plane shear stresses. Increasing the applied out-of-plane shear stresses reduced the membrane shear strength of the elements. The effect of applying out-of-plane shear stresses on the in-plane shear strength was represented by modifying the softening coefficient in the compression stress strain curve of concrete struts. The modified model was able to capture the behavior and the ultimate capacity of the tested elements. The effect of the in-plane shear reinforcement ratio on the interaction between in-plane and out-of-plane shear stresses was evaluated. The model was implemented in the Finite Element package FEAP and was used to predict the ultimate capacity of many structures subjected to a combination of in-plane and out-of-plane shear stresses. The results of the analytical model were used to develop simplified design equations for members subjected to bi-directional shear loads

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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).

  12. 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.

  13. 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.

  14. 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

  15. 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

  16. 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.

  17. Drag and distribution measurements of single-element fuel injectors for supersonic combustors

    NASA Technical Reports Server (NTRS)

    Povinelli, L. A.

    1974-01-01

    The drag caused by several vortex generating fuel injectors for scramjet combustors was measured in a Mach 2 to 3.5 airstream. Injector drag was found to be strongly dependent on injector thickness ratio. The distribution of helium injected into the stream was measured both in the near field and the far field of the injectors for a variety of pressure ratios. The far field results differed appreciably from measurements in the near field. Injection pressure ratio was found to profoundly influence the penetration. One of the aerowing configurations tested yielded low drag consistent with desirable penetration and spreading characteristics.

  18. 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.

  19. 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.

  20. Studies of transient behavior of proton exchange membrane fuel cells (PEMFC)

    NASA Astrophysics Data System (ADS)

    Kim, Sunhoe

    The Proton Exchange Membrane Fuel Cell (PEMFC) is a technology with growing interest. The PEMFC is the most fascinating among other kinds of fuel cells for its high power density and zero emission pure water products. The use of PEMFCs will expose it to transient conditions. For instance, acceleration or deceleration in vehicle applications and turning on or off dishwashers in stationary applications may cause transient conditions of operation of PEMFCs. This dissertation presents experimental data that may be used to understand PEMFC behavior during these transients and these data may be used to verify the numerical simulations and models of PEMFC designs. The electrical load was changed with fixed inlet flowrates for the anode and cathode, and this caused hydrogen and air, stoichiometries to change. The transient experiments showed conditions and stoichiometric changes that gave the overshoot and undershoot behaviors. Data are presented to show the effects of voltage changes on the current response with four different cases of stoichiometry changes: from excess to normal, from normal to excess, from normal to starved, and from starved to normal. An overshoot behavior was observed when the cell stoichiometry changed from normal to starved condition. With a triple path flow field this overshoot was followed by an undershoot and this second order behavior is a result of, in this case, the air flowing back into the cell at the end of anode side to balance pressure. We named these phenomena as "vacuum effects" when the current density shows "undershoot" after "overshoot" behavior. For other conditions an undershoot behavior was observed when the voltage changed to cause a change from starved to normal conditions. In contrast, only exponential first order behavior was observed for voltage changes between excess and normal conditions. Various cell voltage ranges and change rates are presented to compare the overshoot and undershoot behaviors. Experiments were performed

  1. 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.

  2. Comparison of effect of insulating blockages on metal and oxide fuel elements

    SciTech Connect

    Tilbrook, R.W.; Dever, D.J.

    1988-01-01

    The safety philosophy of the new liquid metal reactor (LMR) plant designs is oriented towards inherent protection against loss of coolable geometry and other entries to core disruption. On potential entry is via propagation of local faults. Within this category is a wide range of initiators which each require assessment of their probability and consequences in order to determine their contribution to plant risk. Local faults include those initiators which cause local power/flow disturbances restricted either to a single subassembly or to a local region of the bundle. The concern is that these localized initiators may start a sequence of events in which fuel failure may propagate first within a subassembly envelope and finally cause loss of coolable geometry in adjacent. This document discusses these scenarios. 3 refs., 1 fig.

  3. Comparative Finite Element Analysis of the Stress-Strain States in Three Different Bonded Solid Oxide Fuel Cell Seal Designs

    SciTech Connect

    Weil, K. Scott; Koeppel, Brian J.

    2008-05-15

    One of the critical issues in designing and fabricating a high performance planar solid oxide fuel cell (pSOFC) stack is the development of the appropriate materials and techniques for hermetically sealing the metal and ceramic components. We are currently developing a foil-based approach that appears to offer good hermeticity and mechanical integrity, while minimizing the generation of interfacial stresses in either of the joint substrate materials, particulary the ceramic cell. Prior experimental work conducted on small-scale samples demonstrated the viability of the concept. Here we present recent results from computational analyses undertaken to investigate potential issues associated with scaling up the seal to full-scale pSOFC stack dimensions/geometry. Here we employ finite element modeling to assess the potential thermal cycling performance of this design, specifically as it pertains to sealing components with vastly different thermal expansion properties.

  4. Metal-Element Compounds of Titanium, Zirconium, and Hafnium as Pyrotechnic Fuels

    DTIC Science & Technology

    2015-05-04

    First, the metal-element compounds have negative enthalpies of formation. Second, production of additional reaction products in the liquid and gas...adiabatic reaction temperatures and the thermodynamic products at those temperatures. 3. RESULTS The results of the experimental ignition tests...for a short time. Several examples are presented in Figures 1a-d. Balanced stoichiometries for simple binary combustion reactions may be calculated

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Behavior of transplutonium elements on thin-layer silica gel using di-2-ethylhexylphosphoric acid

    SciTech Connect

    Molochnikova, N.P.; Myasoedov, B.F.

    1995-03-01

    The behavior of transplutonium elements (TPE) on thin-layer silica gel on domestic sorbphil thin-layer chromatography (TLC) plates is investigated using di-2-ethylhexylphosphoric acid (D2EHPA) solutions in HNO{sub 3} of various concentrations. Conditions are found for separating Bk(IV) from other TPE in solutions with [HNO{sub 3}] > 5 M and also for separating Bk(III) and Ce. Solutions of D2EHPA as the mobile phase are recommended for TLC identification of tracer amounts of different oxidation states Am in HNO{sub 3}.

  11. 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.

  12. Superheated steam pyrolysis of biomass elemental components and Sugi (Japanese cedar) for fuels and chemicals.

    PubMed

    Sagehashi, Masaki; Miyasaka, Noritaka; Shishido, Hiromu; Sakoda, Akiyoshi

    2006-07-01

    To develop a novel noncatalytic biomass refinery process that can be used as a portable process, superheated steam pyrolysis was investigated to produce both carbonized solid fuels and chemicals using a large-scale reactor. Individual biomass components and native biomass (Sugi, Japanese cedar) were pyrolyzed. Between 150 and 400 degrees C, the vaporizing fractions of cellulose, xylan, and kraft lignin were summarized using a numerical model. Cellulose was converted to glycolaldehyde, furfural, 5-hydroxymethyl furfural and levoglucosan, whereas xylan was converted to glycolaldehyde, furfural, and acetic acid. Kraft lignin produced a slight yield of phenol and guaiacol. The total vaporization fraction of Sugi and its vaporizing rate were explained sufficiently using a numerical model based on the weighted average of the vaporizing properties of the individual components. However, the yields of phenol, guaiacol, and acetic acid were underestimated, while the yields of furfurals and levoglucosan were overestimated. Possible synergetic effects among chemicals in the superheated steam pyrolysis of native biomass were also discussed.

  13. Report of the CCQM-K123: trace elements in biodiesel fuel

    NASA Astrophysics Data System (ADS)

    Kuroiwa, T.; Zhu, Y.; Inagaki, K.; Long, S. E.; Christopher, S. J.; Puelles, M.; Borinsky, M.; Hatamleh, N.; Murby, J.; Merrick, J.; White, I.; Saxby, D.; Sena, R. C.; Almeida, M. D.; Vogl, J.; Phukphatthanachai, P.; Fung, W. H.; Yau, H. P.; Okumu, T. O.; Kang'iri, J. N.; Télle, J. A. S.; Campos, E. Z.; Gal&vacute; n, E. C.; Kaewkhomdee, N.; Taebunpakul, S.; Thiengmanee, U.; Yafa, C.; Tokman, N.; Tunç, M.; Can, S. Z.

    2017-01-01

    The CCQM-K123 key comparison was organized by the Inorganic Analysis Working Group (IAWG) of CCQM to assess and document the capabilities of the national metrology institutes (NMIs) or the designated institutes (DIs) to measure the mass fractions of sodium, calcium, potassium, magnesium phosphorous and sulfur in biodiesel fuel (BDF). National Metrology Institute of Japan (NMIJ) and National Institute of Standards and Technology (NIST) acted as the coordinating laboratories. Results were submitted by 11 NMIs and DIs. Most of the participants used inductively coupled plasma-mass spectrometry (ICP-MS), isotope dilution technique with ICP-MS and inductively coupled plasma-optical emission spectrometry (ICP-OES) with microwave acid digestion. Accounting for relative expanded uncertainty, comparability of measurement results for each of Na, Ca, K, Mg and P was successfully demonstrated by the participants. Concerning S, the variation in results between participants, particularly those using IDMS methods was observed. According to the additional evaluation and investigation, the revised results were overlapping between IDMS measurements at the k = 2 level. However, this KC does not support S measurements. Main text To reach the main text of this paper, click on Final Report. Note that this text is that which appears in Appendix B of the BIPM key comparison database kcdb.bipm.org/. The final report has been peer-reviewed and approved for publication by the CCQM, according to the provisions of the CIPM Mutual Recognition Arrangement (CIPM MRA).

  14. 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.

  15. The influence of multiple goals on driving behavior: the case of safety, time saving, and fuel saving.

    PubMed

    Dogan, Ebru; Steg, Linda; Delhomme, Patricia

    2011-09-01

    Due to the innate complexity of the task drivers have to manage multiple goals while driving and the importance of certain goals may vary over time leading to priority being given to different goals depending on the circumstances. This study aimed to investigate drivers' behavioral regulation while managing multiple goals during driving. To do so participants drove on urban and rural roads in a driving simulator while trying to manage fuel saving and time saving goals, besides the safety goals that are always present during driving. A between-subjects design was used with one group of drivers managing two goals (safety and fuel saving) and another group managing three goals (safety, fuel saving, and time saving) while driving. Participants were provided continuous feedback on the fuel saving goal via a meter on the dashboard. The results indicate that even when a fuel saving or time saving goal is salient, safety goals are still given highest priority when interactions with other road users take place and when interacting with a traffic light. Additionally, performance on the fuel saving goal diminished for the group that had to manage fuel saving and time saving together. The theoretical implications for a goal hierarchy in driving tasks and practical implications for eco-driving are discussed.

  16. 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.

  17. 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…

  18. 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.

  19. 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

  20. Thermophysical and Corrosion Characteristics of the Actual and Potential Fuel-Element Jackets of Light-Water Reactors in the Case of an Accident with Coolant Loss

    NASA Astrophysics Data System (ADS)

    Bazyuk, S. S.; Kiselev, D. S.; Kuzma-Kichta, Yu. A.; Mokrushin, A. A.; Parshin, N. Ya.; Popov, E. B.; Soldatkin, D. M.; Fedik, I. I.

    2017-01-01

    Corrosion tests were carried out and experimental data are presented on the characteristics of steam-phase oxidation of fuel-element jackets made from an É110G zirconium alloy and vacuum-melted molybdenum in a temperature range of up to 1400°C at atmospheric pressure. The shut-down cooling characteristics on repeated flooding of model fuel assemblies of large-scale stands with fuel elements simulators having jackets made from various standard materials and those from a promising molybdenum one are compared. It is shown that interaction of steam with zirconium alloy is more intense than with molybdenum. At the present time the data on molybdenum are limited.

  1. 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.

  2. 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

  3. 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.

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

    SciTech Connect

    Shiller, A.M. ); Boyle, E.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 variations in river concentrations is a more likely explanation for this behavior. For some of the 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.

  5. 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

  6. 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.

  7. 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.

  8. A Computational and Experimental Study of Ignition Behavior of Gasoline Surrogate Fuels Under Low-Temperature Combustion Conditions

    NASA Astrophysics Data System (ADS)

    Han, J.; Haworth, D. C.; Kalaskar, V. B.; Boehman, A. L.

    2016-11-01

    One strategy for next-generation engines is low-temperature compression ignition of gasoline. Reaction pathways that are not relevant for high-temperature flame propagation are activated under these conditions, and the ignition behavior of these fuels under low-temperature conditions has not been widely explored. Here the ignition behavior of gasoline and two- and three-component surrogates has been studied experimentally and computationally over a range of operating conditions of interest for low-temperature engine combustion. Experiments were performed in a single-cylinder research engine. For each fuel blend, the critical compression ratio (lowest compression ratio at which the main ignition occurs) was determined over a range of operating conditions, by varying one parameter at a time with all other parameters held fixed. A simplified CFD model that considers detailed chemical kinetics was used to simulate the experiment. The focus of the study is to determine which surrogate fuel mixtures and chemical mechanisms are able to capture the ignition behavior of gasoline under these conditions. For example, different ignition behavior is found for different surrogate mixtures that all have the same Research Octane Number, and it is important to capture this behavior in CFD models.

  9. Dynamic behavior of water within a polymer electrolyte fuel cell membrane at low hydration levels.

    PubMed

    Pivovar, Adam M; Pivovar, Bryan S

    2005-01-20

    Protonic conduction across the membrane of a polymer electrolyte fuel cell is intimately related to the dynamic behavior of water present within the membrane. To further the understanding of water dynamics in these materials, quasielastic neutron scattering (QENS) has been used to investigate the picosecond dynamic behavior of water within a perfluorosulfonated ionomer (PFSI) membrane under increasing hydration levels from dry to saturation. Evaluation of the elastic incoherent structure factor (EISF) reveals an increase in the characteristic length-scale of confinement as the number of water molecules in the membrane increases, tending to an asymptotic value at saturation. The fraction of elastic incoherent scattering observed at high Q over all hydration levels is well fit by a simple model that assumes a single, nondiffusing hydronium ion per membrane sulfonic acid site. The quasielastic component of the fitted data indicates confined dynamic behavior for scattering vectors less than 0.7 A(-1). As such, the dynamic behavior was interpreted using continuous diffusion confined within a sphere at Q < 0.7 A(-1) and random unconstrained jump diffusion at Q > 0.7 A(-1). As the number of water molecules in the membrane increases, the characteristic residence times obtained from both models is reduced. The increased dynamical frequency is further reflected in the diffusion coefficients predicted by both models. Between low hydration (2 H2O/SO3H) and saturation (16 H2O/SO3H), the continuous spherical diffusion coefficient changes from 0.46 +/- 0.12 to 1.04 +/- 0.12 (10(-5) cm2/s) and jump diffusion indicates an increase from 1.21 +/- 0.03 to 2.14 +/- 0.08 (10(-5) cm2/s). Overall, the dynamic behavior of water has been quantified over different length scale regimes, the results of which may be rationalized on the basis of the formation of water clusters in the hydrophilic domain that expand toward an asymptotic upper limit with increased hydration.

  10. 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.

  11. 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

    ... the integrity of completed fuel pins (or rods). This item typically includes equipment for: (i) X-ray... (iii) Gamma-ray scanning of the pins (or rods) to check for correct loading of the fuel pellets inside....

  12. 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

    ... (iii) Gamma-ray scanning of the pins (or rods) to check for correct loading of the fuel pellets inside. ... the integrity of completed fuel pins (or rods). This item typically includes equipment for: (i)...

  13. Consumer behavior towards fuel efficient vehicles. Volume II: consumer sentiments towards fuel efficient vehicles. Final report Sep 77-Feb 80

    SciTech Connect

    Sherman, L.; Manski, C.F.; Ginn, J.R.; Hill, D.H.; Hill, M.S.

    1980-02-01

    To meet the National Highway Traffic Safety Administration's (NHTSA) need for a capability of forecasting consumer response to fuel-efficient motor vehicle designs, a statistical analysis was performed on a longitudinal series of Surveys of Consumer Sentiment, administered by the University of Michigan's Survey Research Center. This volume examines a broad set of questions concerning consumer attitudes towards energy policy in general and motor vehicle-related conservation efforts in particular.

  14. 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.

  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. Lateral Flow Field Behavior Downstream of Mixing Vanes In a Simulated Nuclear Fuel Rod Bundle

    SciTech Connect

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2004-07-01

    To assess the fuel assembly performance of PWR nuclear fuel assemblies, average subchannel flow values are used in design analyses. However, for this highly complex flow, it is known that local conditions around fuel rods vary dependent upon the location of the fuel rod in the fuel assembly and upon the support grid design that maintains the fuel rod pitch. To investigate the local flow in a simulated nuclear fuel rod bundle, a testing technique has been employed to measure the lateral flow field in a 5 x 5 rod bundle. Particle Image Velocimetry was used to measure the lateral flow field downstream of a support grid with mixing vanes for four unique subchannels in the 5 x 5 bundle. The dominant lateral flow structures for each subchannel are compared in this paper including the decay of these flow structures. (authors)

  18. 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

  19. 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.

  20. Homogenized finite element analysis on effective elastoplastic mechanical behaviors of composite with imperfect interfaces.

    PubMed

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

    2014-12-16

    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.

  1. Behavioral pattern of a monopolar passive direct methanol fuel cell stack

    NASA Astrophysics Data System (ADS)

    Kim, Young-Jin; Bae, Byungchan; Scibioh, M. Aulice; Cho, EunAe; Ha, Heung Yong

    A passive, air-breathing, monopolar, liquid feed direct methanol fuel cell (DMFC) stack consisting of six unit cells with no external pump, fan or auxiliary devices to feed the reactants has been designed and fabricated for its possible employment as a portable power source. The configurations of the stack of monopolar passive feed DMFCs are different from those of bipolar active feed DMFCs and therefore its operational characteristics completely vary from the active ones. Our present investigation primarily focuses on understanding the unique behavioral patterns of monopolar stack under the influence of certain operating conditions, such as temperature, methanol concentration and reactants feeding methods. With passive reactants supply, the temperature of the stack and open circuit voltage (OCV) undergo changes over time due to a decrease in concentration of methanol in the reservoir as the reaction proceeds. Variations in performance and temperature of the stack are mainly influenced by the concentration of methanol. Continuous operation of the passive stack is influenced by the supply of methanol rather than air supply or water accumulation at the cathode. The monopolar stack made up of six unit cells exhibits a total power of 1000 mW (37 mW cm -2) with 4 M methanol under ambient conditions.

  2. Influence of cell voltage and current on sulfur poisoning behavior of solid oxide fuel cells

    NASA Astrophysics Data System (ADS)

    Cheng, Zhe; Zha, Shaowu; Liu, Meilin

    The sulfur poisoning behavior of nickel-yttria stabilized zirconia (YSZ) cermet anodes in solid oxide fuel cells (SOFCs) was investigated under both potentiostatic and galvanostatic conditions. While the observed relative drop in cell power output caused by sulfur poisoning decreases as the cell-terminal voltage is lowered potentiostatically (thus more current passing through the cell), it increases as more current is drawn from the cell galvanostatically (thus leading to lower terminal voltage). The apparent contradictory trends in relative performance loss due to sulfur poisoning are explained using a simple equivalent circuit analysis, which was further validated by impedance measurements of cells before and after poisoning by trace amounts of hydrogen sulfide (H 2S) under different conditions. Results suggest that the relative increase in cell internal resistance caused by sulfur poisoning is smaller when more current is drawn from the cell (or the cell-terminal voltage is lowered) under either potentiostatic or galvanostatic conditions. Thus, the increase in anode polarization resistance, not the drop in cell power output, should be used to describe the degree of sulfur poisoning in order to avoid any confusion.

  3. The chemical behavior of the transuranic elements and the barrier function in natural aquifer systems

    SciTech Connect

    Jewett, J.R.

    1997-09-17

    In a geological repository for long-lived radioactive wastes, such as actinides and certain fission products, most of the stored radionuclides remain immobile in the particular geological formation. If any of these could possibly become mobile, only trace concentrations of a few radionuclides would result. Nevertheless, with an inventory in the repository of many tonnes of transuranic elements, the amounts that could disperse cannot be neglected. A critical assessment of the chemical behavior of these nuclides, especially their migration properties in the aquifer system around the repository site, is mandatory for analysis of the long-term safety. The chemistry requited for this includes many geochemical multicomponent reactions that are so far only partially understood and [which] therefore can be quantified only incompletely. A few of these reactions have been discussed in this paper based on present knowledge. If a comprehensive discussion of the subject is impossible because of this [lack of information], then an attempt to emphasize the importance of the predominant geochemical reactions of the transuranic elements in various aquifer systems should be made.

  4. 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.

  5. 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.

  6. Finite element analysis of mechanical behavior of human dysplastic hip joints: a systematic review.

    PubMed

    Vafaeian, B; Zonoobi, D; Mabee, M; Hareendranathan, A R; El-Rich, M; Adeeb, S; Jaremko, J L

    2017-04-01

    Developmental dysplasia of the hip (DDH) is a common condition predisposing to osteoarthritis (OA). Especially since DDH is best identified and treated in infancy before bones ossify, there is surprisingly a near-complete absence of literature examining mechanical behavior of infant dysplastic hips. We sought to identify current practice in finite element modeling (FEM) of DDH, to inform future modeling of infant dysplastic hips. We performed multi-database systematic review using PRISMA criteria. Abstracts (n = 126) fulfilling inclusion criteria were screened for methodological quality, and results were analyzed and summarized for eligible articles (n = 12). The majority of the studies modeled human adult dysplastic hips. Two studies focused on etiology of DDH through simulating mechanobiological growth of prenatal hips; we found no FEM-based studies in infants or children. Finite element models used either patient-specific geometry or idealized average geometry. Diversities in choice of material properties, boundary conditions, and loading scenarios were found in the finite-element models. FEM of adult dysplastic hips demonstrated generally smaller cartilage contact area in dysplastic hips than in normal joints. Contact pressure (CP) may be higher or lower in dysplastic hips depending on joint geometry and mechanical contribution of labrum (Lb). FEM of mechanobiological growth of prenatal hip joints revealed evidence for effects of the joint mechanical environment on formation of coxa valga, asymmetrically shallow acetabulum and malformed femoral head associated with DDH. Future modeling informed by the results of this review may yield valuable insights into optimal treatment of DDH, and into how and why OA develops early in DDH.

  7. Identifying Common Practice Elements to Improve Social, Emotional, and Behavioral Outcomes of Young Children in Early Childhood Classrooms.

    PubMed

    McLeod, Bryce D; Sutherland, Kevin S; Martinez, Ruben G; Conroy, Maureen A; Snyder, Patricia A; Southam-Gerow, Michael A

    2017-02-01

    Educators are increasingly being encouraged to implement evidence-based interventions and practices to address the social, emotional, and behavioral needs of young children who exhibit problem behavior in early childhood settings. Given the nature of social-emotional learning during the early childhood years and the lack of a common set of core evidence-based practices within the early childhood literature, selection of instructional practices that foster positive social, emotional, and behavioral outcomes for children in early childhood settings can be difficult. The purpose of this paper is to report findings from a study designed to identify common practice elements found in comprehensive intervention models (i.e., manualized interventions that include a number of components) or discrete practices (i.e., a specific behavior or action) designed to target social, emotional, and behavioral learning of young children who exhibit problem behavior. We conducted a systematic review of early childhood classroom interventions that had been evaluated in randomized group designs, quasi-experimental designs, and single-case experimental designs. A total of 49 published articles were identified, and an iterative process was used to identify common practice elements. The practice elements were subsequently reviewed by experts in social-emotional and behavioral interventions for young children. Twenty-four practice elements were identified and classified into content (the goal or general principle that guides a practice element) and delivery (the way in which a teacher provides instruction to the child) categories. We discuss implications that the identification of these practice elements found in the early childhood literature has for efforts to implement models and practices.

  8. Postirradiation examination of the HT9 clad fuel test X425 at 2.9% burnup

    SciTech Connect

    Pahl, R G; Beck, W N; Sanecki, J E

    1987-11-01

    The X425 experiment was the first EBR-II subassembly to be irradiated with U-Pu-Zr metallic fuel clad in the HT9 alloy. This report summarizes our initial postirradiation examination of selected elements from X425 at 2.9% peak burnup. Fuel microstructure, swelling behavior, fission gas release, and fuel/clad chemical interaction are discussed.

  9. Human skeletal muscle behavior in vivo: Finite element implementation, experiment, and passive mechanical characterization.

    PubMed

    Clemen, Christof B; Benderoth, Günther E K; Schmidt, Andreas; Hübner, Frank; Vogl, Thomas J; Silber, Gerhard

    2017-01-01

    In this study, useful methods for active human skeletal muscle material parameter determination are provided. First, a straightforward approach to the implementation of a transversely isotropic hyperelastic continuum mechanical material model in an invariant formulation is presented. This procedure is found to be feasible even if the strain energy is formulated in terms of invariants other than those predetermined by the software's requirements. Next, an appropriate experimental setup for the observation of activation-dependent material behavior, corresponding data acquisition, and evaluation is given. Geometry reconstruction based on magnetic resonance imaging of different deformation states is used to generate realistic, subject-specific finite element models of the upper arm. Using the deterministic SIMPLEX optimization strategy, a convenient quasi-static passive-elastic material characterization is pursued; the results of this approach used to characterize the behavior of human biceps in vivo indicate the feasibility of the illustrated methods to identify active material parameters comprising multiple loading modes. A comparison of a contact simulation incorporating the optimized parameters to a reconstructed deformed geometry of an indented upper arm shows the validity of the obtained results regarding deformation scenarios perpendicular to the effective direction of the nonactivated biceps. However, for a valid, activatable, general-purpose material characterization, the material model needs some modifications as well as a multicriteria optimization of the force-displacement data for different loading modes.

  10. Influence of superstructure geometry on the mechanical behavior of zirconia implant abutments: a finite element analysis.

    PubMed

    Geringer, Alexander; Diebels, Stefan; Nothdurft, Frank P

    2014-12-01

    To predict the clinical performance of zirconia abutments, it is crucial to examine the mechanical behavior of different dental implant-abutment connection configurations. The international standard protocol for dynamic fatigue tests of dental implants (ISO 14801) allows comparing these configurations using standardized superstructure geometries. However, from a mechanical point of view, the geometry of clinical crowns causes modified boundary conditions. The purpose of this finite element (FE) study was to evaluate the influence of the superstructure geometry on the maximum stress values of zirconia abutments with a conical implant-abutment connection. Geometry models of the experimental setup described in ISO 14801 were generated using CAD software following the reconstruction of computerized tomography scans from all relevant components. These models served as a basis for an FE simulation. To reduce the numerical complexity of the FE model, the interaction between loading stamp and superstructure geometry was taken into account by defining the boundary conditions with regard to the frictional force. The results of the FE simulations performed on standardized superstructure geometry and anatomically shaped crowns showed a strong influence of the superstructure geometry and related surface orientations on the mechanical behavior of the underlying zirconia abutments. In conclusion, ISO testing of zirconia abutments should be accompanied by load-bearing capacity testing under simulated clinical conditions to predict clinical performance.

  11. Behavioral effects of low-level exposure to elemental Hg among dentists.

    PubMed

    Echeverria, D; Heyer, N J; Martin, M D; Naleway, C A; Woods, J S; Bittner, A C

    1995-01-01

    Exposure thresholds for health effects associated with elemental mercury (Hg degree) exposure were examined by comparing behavioral test scores of 19 exposed (mean urinary Hg = 36 micrograms/l) with those of 20 unexposed dentists. Thirty-six micrograms Hg/l is 7 times greater than the 5 micrograms Hg/l mean level measured in a national sample of dentists. To improve the distinction between recent and cumulative effects, the study also evaluated porphyrin concentrations in urine, which are correlated with renal Hg content (a measure of cumulative body burden). Subjects provided an on-site spot urine sample, were administered a 1-h assessment consisting of a consent form, the Profile of Mood Scales, a symptom and medical questionnaire, and 6 behavioral tests: digit-span, symbol-digit substitution, simple reaction time, the ability to switch between tasks, vocabulary, and the One Hole Test. Multivariate regression techniques were used to evaluate dose-effects controlling for the effects of age, race, gender and alcohol consumption. A dose-effect was considered statistically significant below a p value of 0.05. Significant urinary Hg dose-effects were found for poor mental concentration, emotional lability, somatosensory irritation, and mood scores. Individual tests evaluating cognitive and motor function changed in the expected directions but were not significantly associated with urinary Hg. However, the pooled sum of rank scores for combinations of tests within domains were significantly associated with urinary Hg, providing evidence of subtle preclinical changes in behavior associated with Hg exposure. Coproporphyrin, one of three urinary porphyrins altered by mercury exposure, was significantly associated with deficits in digit span and simple reaction time.(ABSTRACT TRUNCATED AT 250 WORDS)

  12. 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.

  13. The study of wear behaviors on abducted hip joint prostheses by an alternate finite element approach.

    PubMed

    Lin, Yi-Tsung; Wu, James Shih-Shyn; Chen, Jian-Horng

    2016-07-01

    An acetabular cup with larger abduction angles is able to affect the normal function of the cup seriously that may cause early failure of the total hip replacement (THR). Complexity of the finite element (FE) simulation in the wear analysis of the THR is usually concerned with the contact status, the computational effort, and the possible divergence of results, which become more difficult on THRs with larger cup abduction angles. In the study, we propose a FE approach with contact transformation that offers less computational effort. Related procedures, such as Lagrangian Multiplier, partitioned matrix inversion, detection of contact forces, continuity of contact surface, nodal area estimation, etc. are explained in this report. Through the transformed methodology, the computer round-off error is tremendously reduced and the embedded repetitive procedure can be processed precisely and quickly. Here, wear behaviors of THR with various abduction angles are investigated. The most commonly used combination, i.e., metal-on-polyethylene, is adopted in the current study where a cobalt-chromium femoral head is paired with an Ultra High Molecular Weight Polyethylene (UHMWPE) cup. In all illustrations, wear coefficients are estimated by self-averaging strategy with available experimental datum reported elsewhere. The results reveal that the THR with larger abduction angles may produce deeper depth of wear but the volume of wear presents an opposite tendency; these results are comparable with clinical and experimental reports. The current approach can be widely applied easily to fields such as the study of the wear behaviors on ante-version, impingement, and time-dependent behaviors of prostheses etc.

  14. 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.

  15. The importance of sulfur for the behavior of highly-siderophile elements during Earth's differentiation

    NASA Astrophysics Data System (ADS)

    Laurenz, Vera; Rubie, David C.; Frost, Daniel J.; Vogel, Antje K.

    2016-12-01

    The highly siderophile elements (HSEs) are widely used as geochemical tracers for Earth's accretion and core formation history. It is generally considered that core formation strongly depleted the Earth's mantle in HSEs, which were subsequently replenished by a chondritic late veneer. However, open questions remain regarding the origin of suprachondritic Ru/Ir and Pd/Ir ratios that are thought to be characteristic for the primitive upper mantle. In most core-formation models that address the behavior of the HSEs, light elements such as S entering the core have not been taken into account and high P-T experimental data for S-bearing compositions are scarce. Here we present a comprehensive experimental study to investigate the effect of increasing S concentration in the metal on HSE metal-silicate partitioning at 2473 K and 11 GPa. We show that the HSEs become less siderophile with increasing S concentrations in the metal, rendering core-forming metal less efficient in removing the HSEs from the mantle if S is present. Furthermore, we investigated the FeS sulfide-silicate partitioning of the HSEs as a function of pressure (7-21 GPa) and temperature (2373-2673 K). The sulfide-silicate partition coefficient for Pt increases strongly with P, whereas those for Pd, Ru and Ir all decrease. The combined effect is such that above ∼20 GPa Ru becomes less chalcophile than Pt, which is opposite to their behavior in the metal-silicate system where Ru is always more siderophile than Pt. The newly determined experimental results are used in a simple 2-stage core formation model that takes into account the effect of S on the behavior of the HSEs during core formation. Results of this model show that segregation of a sulfide liquid to the core from a mantle with substantial HSE concentrations plays a key role in reproducing Earth's mantle HSE abundances. As Ru and Pd are less chalcophile than Pt and Ir at high P-T, some Ru and Pd remain in the mantle after sulfide segregation

  16. 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

  17. Fuel Surrogate Physical Property Effects on Direct Injection Spray and Ignition Behavior

    DTIC Science & Technology

    2015-09-01

    jet was steady. As shown in Figure 2, the model was capable of capturing the general radial and axial fuel vapor distribution trends. At 25 mm...UNCLASSIFIED 1 Introduction Typical hydrocarbon fuels used in internal combustion engines, such as gasoline, diesel, or jet fuel, are composed of hundreds...to thousands of hydrocarbon (HC) species. Such a large number of species in high fidelity Computational Fluid Dynamics (CFD) with detailed chemistry

  18. Effect [of] co-combustion of sewage sludge and biomass on combustion behavior and emissions in pulverized fuel systems

    SciTech Connect

    Spliethoff, H.; Hein, K.R.G.

    1999-07-01

    Biomass not only has a considerable potential as an additional fuel source but also shows a reasonable cost level in comparison to other renewable energies. The practicable fuel types are both residual material from forestry and agriculture, such as wood or straw, and especially cultivated reproducible feedstock such as Miscanthus Sinensis, whole cereal plants, poplars, or willows. Besides as single fuel, it is also considered to be sensible to utilize biomass in co-combustion in existing firing systems, such as pc-fired power stations. Biomass or sewage sludge utilized as additional fuel in coal combustion systems has consequences on combustion behavior, emissions, corrosion and residual matter. The effects of burning sewage sludge and agricultural residuals such as straw and manure as well as specially grown energy plants in combination with coal were studied in a 0.5 MW pulverized fuel test facility and a 20 kW electrically heated combustor. A major aspect of the investigations had been the required preparation and milling of the additional fuels. The investigations showed that in co-combustion of straw with coal, a grinding of 6 mm and finer is sufficient. The definitely coarser milling degree of biomass delays combustion and is observable by in-flame measurements. The investigations reveal that biomass addition has a positive effect on emissions. Since biomass in most cases contains considerably less sulphur than coal, an increasing biomass share in the thermal output makes the SO{sub 2} emissions decrease proportionally. In addition, SO{sub 2} can partly be captured in the ash by the alkaline-earth fractions of the biomass ash. As for sewage sludge, the emissions of SO{sub 2} correlate with the sulphur content of the fuel and, hence, rise with an increasing share of this biomass. Independently from the type, biomass shows a considerably stronger release of volatile matter. This latter fact may have a positive impact on NOx emissions when NOx

  19. 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.

  20. Experimental Study of Low Temperature Behavior of Aviation Turbine Fuels in a Wing Tank Model

    NASA Technical Reports Server (NTRS)

    Stockemer, Francis J.

    1979-01-01

    An experimental investigation was performed to study aircraft fuels at low temperatures near the freezing point. The objective was an improved understanding of the flowability and pumpability of the fuels under conditions encoutered during cold weather flight of a long range commercial aircraft. The test tank simulated a section of an outer wing tank and was chilled on the upper and lower surfaces. Fuels included commercial Jet A and Diesel D-2; JP-5 from oil shale; and Jet A, intermediate freeze point, and D-2 fuels derived from selected paraffinic and naphthenic crudes. A pour point depressant was tested.

  1. 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

  2. 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

  3. 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.

  4. Influence of Group IV and V Alloying Elements on the Microstructure Engineering and Deformation Behavior in Tantalum Carbides

    DTIC Science & Technology

    2015-04-06

    stable for a range of stoichiometry allowing a large amount of carbon loss. For the group VB compounds, the FCC structure of the metal atoms is...AFRL-OSR-VA-TR-2015-0092 Influence of Group IV and V Alloying Elements on the Microstructure Engineering GREGORY THOMPSON UNIVERSITY OF ALABAMA Final...31/03/2015 4. TITLE AND SUBTITLE Influence of Group IV and V Alloying Elements on the Microstructure Engineering and Deformation Behavior in

  5. 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.

  6. 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.

  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

    2016-09-06

    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.

  8. 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

  9. Complex Source and Radiation Behaviors of Small Elements of Linear and Matrix Flexible Ultrasonic Phased-Array Transducers

    NASA Astrophysics Data System (ADS)

    Amory, V.; Lhémery, A.

    2008-02-01

    Inspection of irregular components is problematical: maladjustment of transducer shoes to surfaces causes aberrations. Flexible phased-arrays (FPAs) designed at CEA LIST to maximize contact are driven by adapted delay laws to compensate for irregularities. Optimizing FPA requires simulation tools. The behavior of one element computed by FEM is observed at the surface and its radiation experimentally validated. Efforts for one element prevent from simulating a FPA by FEM. A model is proposed where each element behaves as nonuniform source of stresses. Exact and asymptotic formulas for Lamb problem are used as convolution kernels for longitudinal, transverse and head waves; the latter is of primary importance for angle-T-beam inspections.

  10. Rare earth element behavior during groundwater-seawater mixing along the Kona Coast of Hawaii

    NASA Astrophysics Data System (ADS)

    Johannesson, Karen H.; Palmore, C. Dianne; Fackrell, Joseph; Prouty, Nancy G.; Swarzenski, Peter W.; Chevis, Darren A.; Telfeyan, Katherine; White, Christopher D.; Burdige, David J.

    2017-02-01

    Groundwater and seawater samples were collected from nearshore wells and offshore along the Kona Coast of the Big Island of Hawaii to investigate rare earth element (REE) behavior in local subterranean estuaries. Previous investigations showed that submarine groundwater discharge (SGD) is the predominant flux of terrestrial waters to the coastal ocean along the arid Kona Coast of Hawaii. Groundwater and seawater samples were filtered through 0.45 μm and 0.02 μm pore-size filters to evaluate the importance of colloidal and soluble (i.e., truly dissolved ionic species and/or low molecular weight [LMW] colloids) fractions of the REEs in the local subterranean estuaries. Mixing experiments using groundwater collected immediately down gradient from a wastewater treatment facility (WWTF) proximal to the Kaloko-Hanokohau National Historic Park, and more "pristine" groundwater from a well constructed in a lava tube at Kiholo Bay, were conducted with local seawater to study the effect of solution composition (i.e., pH, salinity) on the concentrations and fractionation behavior of the REEs as groundwater mixes with seawater in Kona Coast subterranean estuaries. The mixed waters were also filtered through 0.45 or 0.02 μm filters to ascertain the behavior of colloidal and soluble fractions of the REEs across the salinity gradient in each mixing experiment. Concentrations of the REEs were statistically identical (two-tailed Student t-test, 95% confidence) between the sequentially filtered sample aliquots, indicating that the REEs occur as dissolved ionic species and/or LMW colloids in Kona Coast groundwaters. The mixing experiments revealed that the REEs are released to solution from suspended particles or colloids when Kona Coast groundwater waters mix with local seawater. The order of release that accompanies increasing pH and salinity follows light REE (LREE) > middle REE (MREE) > heavy REE (HREE). Release of REEs in the mixing experiments is driven by decreases in the

  11. Rare earth element behavior during groundwater – seawater mixing along the Kona Coast of Hawaii

    USGS Publications Warehouse

    Johannesson, Karen H.; Palmore, C. Dianne; Fackrell, Joseph; Prouty, Nancy G.; Swarzenski, Peter W.; Chevis, Darren A.; Telfeyan, Katherine; White, Christopher D.; Burdige, David J.

    2017-01-01

    Groundwater and seawater samples were collected from nearshore wells and offshore along the Kona Coast of the Big Island of Hawaii to investigate rare earth element (REE) behavior in local subterranean estuaries. Previous investigations showed that submarine groundwater discharge (SGD) is the predominant flux of terrestrial waters to the coastal ocean along the arid Kona Coast of Hawaii. Groundwater and seawater samples were filtered through 0.45 μm and 0.02 μm pore-size filters to evaluate the importance of colloidal and soluble (i.e., truly dissolved ionic species and/or low molecular weight [LMW] colloids) fractions of the REEs in the local subterranean estuaries. Mixing experiments using groundwater collected immediately down gradient from a wastewater treatment facility (WWTF) proximal to the Kaloko-Hanokohau National Historic Park, and more “pristine” groundwater from a well constructed in a lava tube at Kiholo Bay, were mixed with local seawater to study the effect of solution composition (i.e., pH, salinity) on the concentrations and fractionation behavior of the REEs as groundwater mixes with seawater in Kona Coast subterranean estuaries. The mixed waters were also filtered through 0.45 or 0.02 μm filters to ascertain the behavior of colloidal and soluble fractions of the REEs across the salinity gradient in each mixing experiment. Concentrations of the REEs were statistically identical (two-tailed Student t-test, 95% confidence) between the sequentially filtered sample aliquots, indicating that the REEs occur as dissolved ionic species and/or LMW colloids in Kona Coast groundwaters. The mixing experiments revealed that the REEs are released to solution from suspended particles or colloids when Kona Coast groundwater waters mix with local seawater. The order of release that accompanies increasing pH and salinity follows light REE (LREE) > middle REE (MREE) > heavy REE (HREE). Release of REEs in the mixing experiments is driven by decreases in the

  12. Scoping studies of vapor behavior during a severe accident in a metal-fueled reactor

    SciTech Connect

    Spencer, B.W.; Marchaterre, J.F.

    1985-04-15

    Scoping calculations have been performed examining the consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel. The principal gas and vapor species released are shown to be Xe, Cs,and bond sodium contained within the fuel porosity. Fuel vapor pressure is insignificant, and there is no energetic fuel-coolant interaction for the conditions considered. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core (although reactor-material experiments are needed to confirm these high condensation rates). If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the implication is that the ability of vapor expansion to perform appreciable work on the system is largely eliminated. Furthermore, the ability of an expanding vapor bubble to transport fuel and fission product species to the cover gas region where they may be released to the containment is also largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool.

  13. 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.

  14. Analysis of the Dynamic Behavior of the Inner Hair Cell Stereocilia by the Finite Element Method

    NASA Astrophysics Data System (ADS)

    Matsui, Toshihiro; Nakajima, Chihiro; Yamamoto, Yuichi; Andoh, Masayoshi; Iida, Koji; Murakoshi, Michio; Kumano, Shun; Wada, Hiroshi

    The motion of the inner hair cell (IHC) stereocilia, which results in tension in the tip links connected to mechanically gated ion channels, mediates the auditory transduction process. However, it is difficult to directly observe the motion of the stereocilia because of their minute dimensions and complex structure. In this study, to investigate such motion, a finite element method model of the tall, middle and short IHC stereocilia, including the tip and lateral links extending between the stereocilia, was constructed. By applying an analytically estimated fluid force caused by a stimulus of 60dB SPL at 500Hz to the model, the dynamic behavior of the stereocilia was analyzed. Numerical results showed that the stereocilia moved in phase and that the maximum tensions of 2.5fN and 2.1fN occurred in the tip link connecting the tall and middle stereocilia and in the tip link connecting the middle and short stereocilia, respectively. By contrast, under the condition in which the lateral links were removed, maximum tension in the former increased to 11.6fN, while that in the latter only increased to 2.3fN. It was therefore suggested that the lateral links protect the MET channels located at taller stereocilia against large stimuli and subject the channels located in the same IHC to forces of similar size.

  15. 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.

  16. The behavior of rare earth elements in naturally and anthropogenically acidified waters

    USGS Publications Warehouse

    Wood, Scott A.; Gammons, Christopher H.; Parker, Stephen R.

    2006-01-01

    In this paper, the behavior of rare earth elements (REE) in a watershed impacted by acid-mine drainage (Fisher Creek, Montana) is compared to that in a volcanically acidified watershed (Rio Agrio and Lake Caviahue, Argentina). The REE behave conservatively in acidic waters with pH values less than approximately 5.5. However, above pH 5.5, REE concentrations are controlled by adsorption onto or co-precipitation with a variety of Fe or Al oxyhydroxides. The heavy REE partition to a greater extent into the solid phase than the light REE as pH rises above 6. Concentrations of REE exhibit diel (24-h) cycling in waters that were initially acidic, but have become neutralized downstream. In Fisher Creek, at the most downstream sampling station investigated (pH 6.8), concentrations of dissolved REE were 190–840% higher in the early morning versus the late afternoon. This cycling can be related to temperature-dependent, cyclic adsorption–desorption of REE onto hydrous ferric or aluminum oxide or both. Similar but gentler diel cycling of the REE was found at Rio Agrio. The existence of such cycling has important ramifications for the study of REE in natural waters.

  17. [Fire behavior of Mongolian oak leaves fuel-bed under no-wind and zero-slope conditions. I. Factors affecting fire spread rate and modeling].

    PubMed

    Jin, Sen; Liu, Bo-Fei; Di, Xue-Ying; Chu, Teng-Fei; Zhang, Ji-Li

    2012-01-01

    Aimed to understand the fire behavior of Mongolian oak leaves fuel-bed under field condition, the leaves of a secondary Mongolian oak forest in Northeast Forestry University experimental forest farm were collected and brought into laboratory to construct fuel-beds with varied loading, height, and moisture content, and a total of 100 experimental fires were burned under no-wind and zero-slope conditions. It was observed that the fire spread rate of the fuel-beds was less than 0.5 m x min(-1). Fuel-bed loading, height, and moisture contents all had significant effects on the fire spread rate. The effect of fuel-bed moisture content on the fire spread had no significant correlations with fuel-bed loading and height, but the effect of fuel-bed height was related to the fuel-bed loading. The packing ratio of fuel-beds had less effect on the fire spread rate. Taking the fuel-bed loading, height, and moisture content as predictive variables, a prediction model for the fire spread rate of Mongolian oak leaves fuel-bed was established, which could explain 83% of the variance of the fire spread rate, with a mean absolute error 0.04 m x min(-1) and a mean relative error less than 17%.

  18. Analysis of the behavior and degradation in proton exchange membrane fuel cells with a dead-ended anode

    NASA Astrophysics Data System (ADS)

    Yu, Jianliang; Jiang, Zuwei; Hou, Ming; Liang, Dong; Xiao, Yu; Dou, Meiling; Shao, Zhigang; Yi, Baolian

    2014-01-01

    Proton exchange membrane fuel cells (PEMFCs) with a dead-ended anode (DEA) can obtain high hydrogen utilization by a comparatively simple system. Nevertheless, the accumulation of the nitrogen and the water in the anode channels can lead to a local fuel starvation, which degrades the performance and durability of PEMFCs. In this paper, the behaviors of PEMFCs with a DEA are explored experimentally by detecting the current distribution and the local potentials. The results indicate that the current distribution is uneven during the DEA operation. The local current firstly decreases at the region near the anode outlet, and then extends to the inlet region along the channels with time. The complete fuel starvation near the anode outlet leads to a high local potential and carbon corrosion on the cathode side. The SEM images of the cathode electrode reveal that the significant thickness reduction and the collapse of the electrode's porous structure happen in the cathode catalyst layer, leading to the irreversible decline of the performance. The comparison of the experiments with different oxidants and fuels reveals that the nitrogen crossover from cathode to anode is the dominant factor on the performance decline under the DEA operations.

  19. Comparison of degradation behaviors for open-ended and closed proton exchange membrane fuel cells during startup and shutdown cycles

    NASA Astrophysics Data System (ADS)

    Yu, Yi; Tu, Zhengkai; Zhang, Haining; Zhan, Zhigang; Pan, Mu

    As catalyst support materials, the oxidation of carbon materials is considered one of the major factors for performance decay during the startup and shutdown process of proton exchange membrane fuel cells, which must be mitigated to achieve acceptable durability. In this paper, the effect of cathode exhaust conditions on the degradation behaviors of fuel cells is investigated using two single cells named the open-ended and closed cells. The cathode inlet pressure during the introduction of the dummy load is an important factor in analyzing the performance decay of membrane electrode assemblies under different conditions. Electrochemical techniques, including the measurement of polarization curves, cyclic and linear sweep voltammetry, and cross-sectional scanning electron microscopy of tested membrane electrode assemblies, are employed to evaluate the performance decay of fuel cells. The results show that a closed cathode exhaust valve during the introduction of the dummy load would significantly alleviate both the performance decay and the decrease in the electrochemically active surface area, resulting in an improvement in fuel cell durability. No significant deterioration of the membranes is observed for both the open-ended and the closed cells during frequent startup and shutdown processes.

  20. Verification of the BISON fuel performance code

    SciTech Connect

    D. M. Perez; R. J. Gardner; J. D. Hales; S. R. Novascone; G. Pastore; B. W. Spencer; R. L. Williamson

    2014-09-01

    BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Labo- ratory (USA) since 2009. The code is applicable to both steady and transient fuel behavior and is used to analyze 1D spherical, 2D axisymmetric, or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods and other well known fuel performance codes. Results from several assessment cases are reported, with emphasis on fuel centerline temperatures at various stages of fuel life, fission gas release, and clad deformation during pellet clad mechanical interaction (PCMI). BISON comparisons to fuel centerline temperature measurements are very good at beginning of life and reasonable at high burnup. Although limited to date, fission gas release comparisons are very good. Comparisons of rod diameter following significant power ramping are also good and demonstrate BISON’s unique ability to model discrete pellet behavior and accurately predict clad ridging from PCMI.

  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

    2016-07-05

    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.

  2. Consumer behavior towards fuel efficient vehicles. Volume IV: operating instructions and program documentation for the CS vehicle choice simulation model. Final report, October 1977-September 1980

    SciTech Connect

    Ginn, J.R.; Berkovec, J.A.

    1980-09-01

    The report assesses consumer behavior towards fuel-efficient vehicles designed to meet recently mandated federal fuel economy standards. The study involves a comprehensive evaluation of existing nationwide survey data as well as the development of a major new econometric forecasting model of household vehicle type choice. As a result, the report describes both an assessment of consumers' current reported sentiments toward fuel-efficient vehicles and insights into expected future changes in household vehicle purchase behavior in response to changes in vehicle designs and prices, demographics, and the energy environment.

  3. 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.

  4. 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.

  5. [Behavior of fuel hot particles in the body of cows at oral intake].

    PubMed

    Kashparov, V A; Lazarev, N M; Ioshchenko, V I

    1997-01-01

    It was studied the behaviour of fuel hot particles (analogous to Chernobyl) in gastrointestinal tract of cows. The values of caesium and strontium radionuclides transfer to the cows organism and its transition parameters to milk after the single per oral intake to the organism of animals are estimated. It is shown, that the biological simplicity of radionuclides in the fuel hot particles at two parameters lower, than the same radionuclides in washed phases.

  6. A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel

    SciTech Connect

    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 selfmated 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.

  7. 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.

  8. Premixed ignition behavior of alternative diesel fuel-relevant compounds in a motored engine experiment

    SciTech Connect

    Szybist, James P.; Boehman, Andre L.; Haworth, Daniel C.; Koga, Hibiki

    2007-04-15

    A motored engine study using premixed charges of fuel and air at a wide range of diesel-relevant equivalence ratios was performed to investigate autoignition differences among surrogates for conventional diesel fuel, gas-to-liquid (GTL) diesel fuel, and biodiesel, as well as n-heptane. Experiments were performed by delivering a premixed charge of vaporized fuel and air and increasing the compression ratio in a stepwise manner to increase the extent of reaction while monitoring the exhaust composition via Fourier transform infrared (FTIR) spectrometry and collecting condensable exhaust gas for subsequent gas chromatography/mass spectrometry (GC/MS) analysis. Each fuel demonstrated a two-stage ignition process, with a low-temperature heat release (LTHR) event followed by the main combustion, or high-temperature heat release (HTHR). Among the three diesel-relevant fuels, the magnitude of LTHR was highest for GTL diesel, followed by methyl decanoate, and conventional diesel fuel last. FTIR analysis of the exhaust for n-heptane, the conventional diesel surrogate, and the GTL diesel surrogate revealed that LTHR produces high concentrations of aldehydes and CO while producing only negligible amounts of CO{sub 2}. Methyl decanoate differed from the other two-stage ignition fuels only in that there were significant amounts of CO{sub 2} produced during LTHR; this was the result of decarboxylation of the ester group, not the result of oxidation. GC/MS analysis of LTHR exhaust condensate for n-heptane revealed high concentrations of 2,5-heptanedione, a di-ketone that can be closely tied to species in existing autoignition models for n-heptane. GC/MS analysis of the LTHR condensate for conventional diesel fuel and GTL diesel fuel revealed a series of high molecular weight aldehydes and ketones, which were expected, as well as a series of organic acids, which are not commonly reported as products of combustion. The GC/MS analysis of the methyl decanoate exhaust condensate

  9. Static simulation and analyses of mower's ROPS behavior in a finite element model.

    PubMed

    Wang, X; Ayers, P; Womac, A R

    2009-10-01

    The goal of this research was to numerically predict the maximum lateral force acting on a mower rollover protective structure (ROPS) and the energy absorbed by the ROPS during a lateral continuous roll. A finite element (FE) model of the ROPS was developed using elastic and plastic theories including nonlinear relationships between stresses and strains in the plastic deformation range. Model validation was performed using field measurements of ROPS behavior in a lateral continuous roll on a purpose-designed test slope. Field tests determined the maximum deformation of the ROPS of a 900 kg John Deere F925 mower with a 183 cm (72 in.) mowing deck during an actual lateral roll on a pad and on soil. In the FE model, lateral force was gradually added to the ROPS until the field-measured maximum deformation was achieved. The results from the FE analysis indicated that the top corners of the ROPS enter slightly into the plastic deformation region. Maximum lateral forces acting on the ROPS during the simulated impact with the pad and soil were 19650 N and 22850 N, respectively. The FE model predicted that the energy absorbed by the ROPS (643 J) in the lateral roll test on the pad was less than the static test requirements (1575 J) of Organization for Economic Development (OECD) Code 6. In addition, the energy absorbed by the ROPS (1813 J) in the test on the soil met the static test requirements (1575 J). Both the FE model and the field test results indicated that the deformed ROPS of the F925 mower with deck did not intrude into the occupant clearance zone during the lateral continuous or non-continuous roll.

  10. Thermal breeder fuel enrichment zoning

    DOEpatents

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

    1992-01-01

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

  11. 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.

  12. 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.

  13. Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    SciTech Connect

    Brower, Jeffrey O.; Glazoff, Michael V.; Eiden, Thomas J.; Rezvoi, Aleksey V.

    2016-08-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping

  14. Analytical approximations for the long-term decay behavior of spent fuel and high-level waste

    SciTech Connect

    Malbrain, C.M.; Deutch, J.M.; Lester, R.K.

    1982-05-01

    Simple analytical approximations are presented that describe the radioactivity and radiogenic decay heat behavior of high-level wastes (HLWs) from various nuclear fuel cycles during the first 100,000 years of waste life. The correlations are based on detailed computations of HLW properties carried out with the isotope generation and depletion code ORIGEN 2. The ambiguities encountered in using simple comparisons of the hazards posed by HLWs and naturally occurring mineral deposits to establish the longevity requirements for geologic waste disposal schemes are discussed.

  15. Oxidation and reduction behaviors of a prototypic MgO-PuO2-x inert matrix fuel

    NASA Astrophysics Data System (ADS)

    Miwa, Shuhei; Osaka, Masahiko

    2017-04-01

    Oxidation and reduction behaviors of prototypic MgO-based inert matrix fuels (IMFs) containing PuO2-x were experimentally investigated by means of thermogravimetry. The oxidation and reduction kinetics of the MgO-PuO2-x specimen were determined. The oxidation and reduction rates of the MgO-PuO2-x were found to be low compared with those of PuO2-x. It is note that the changes in O/Pu ratios of MgO-PuO2-x from stoichiometry were smaller than those of PuO2-x at high oxygen partial pressure.

  16. 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.

  17. 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.

  18. Determination of trace elements by inductively coupled plasma mass spectrometry of biomass and fuel oil reference materials using milligram sample sizes.

    PubMed

    Lachas, H; Richaud, R; Herod, A A; Dugwell, D R; Kandiyoti, R

    2000-01-01

    Most of the analytical techniques used to quantify elements associated with solid samples suffer from high detection limits and cannot be used for trace elements in biomass samples, particularly when only 20 mg are available for analysis. Inductively coupled plasma mass spectrometry (ICP-MS) can achieve detection limits of parts-per-trillion with liquid sample introduction by solution nebulisation. This technique was therefore tested with two standard biomass reference materials: oriental tobacco leaves and cabbage leaves. Two preparations successfully used on coal standards were used to digest the solid samples: a total digestion method (wet ashing digestion) and a partial leaching (microwave extraction). The concentrations of up to seventeen elements (As, Ba, Be, Cd, Co, Cr, Cu, Ga, Mn, Mo, Ni, Pb, Sb, Se, Sn, V and Zn) were measured after the two preparations. The accuracy and sensitivity of the measurements improved when the dilution factor decreased from 5000 to 1000 and to 500. Since the proportion of mineral matter in biomass samples is small (5%), the microwave digestion extracted elements that are generally not completely extracted from coal samples (e.g. Sb). However, some trace element concentrations were below the limit of quantification after microwave extraction, even with a reduced dilution factor (As, Se and Mo) and could not be quantified. A fuel oil was also digested. The trace element concentrations were very low (between 28 and 0.1 microgram g(-1)) but acceptable results were obtained by applying a dilution factor of 100. Only six elements in the fuel oil (As, Ba, Co, Ni, Se and V) had certified or indicated values. Factors affecting the accuracy and sensitivity of the analyses are discussed. The reproducibility of analysis of the tobacco leaf standard was checked over a period of nine months by both digestion methods. The wet ashing method gave acceptable reproducibility for Ba, Cd, Co, Cu, Ga, Mn, Mo, Ni, Pb, V and Zn but poor precision for Cr

  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. 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.

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

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... pins (or rods); (3) Automatic test and inspection stations especially designed or prepared for checking the integrity of completed fuel pins (or rods). This item typically includes equipment for: (i) X-ray examination of pin (or rod) end cap welds; (ii) Helium leak detection from pressurized pins (or rods);...

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

    EPA Science Inventory

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

  3. Behavior of trace and companion elements of ULC-IF steel grades during RH-treatment

    SciTech Connect

    Jungreithmeier, A.; Viertauer, A.; Presslinger, H.

    1996-12-31

    A large number of metallurgical reactions are caused by lowering the partial pressure during vacuum treatment. One of these reactions is the volatilization of elements with high vapor pressure. The concentration of trace and companion elements during RH-treatment mostly changes because of cooling scrap, deoxidation agents and ferro-alloy additions, slag/metal reactions, vaporization and also because of reactions with the RH-vessel lining. These changes in the concentration of trace and companion elements during RH-treatment are exemplified for ULC-IF (ultra low carbon--interstitial free) steel grades. The elements which are considered are chromium, nickel, molybdenum, copper, vanadium, tin, zinc, lead, phosphorus, sulfur and nitrogen. Calculations of the theoretical equilibrium solubility using thermodynamic data--in dependence of pressure and temperature--correspond well with the values obtained during steel production operations. 67 refs.

  4. Consumer behavior towards fuel efficient vehicles. Volume II. Final report Oct 77-Mar 81

    SciTech Connect

    Heisler, J.T.; Groeneman, S.

    1981-03-01

    The Energy Policy and Conservation Act of 1975 requires fuel efficiency standards through Model Year 1985 and requires that economic practicability (e.g., marketability, consumer impact) be considered in miles per gallon formulations. The purpose of this research is to develop information on likely market response to various vehicle design and performance changes to minimize the possibility of market rejection. The report is based on focus group discussions and consumer experiments with drivers in different vehicle size classes. Analysis focuses on the believability of the energy crisis and concern about its effects, the acceptability of possible fuel economy options, consumer preferences with respect to vehicles embodying selected design and engineering changes, and predicted purchase and usage patterns under different future scenarios.

  5. Consumer behavior towards fuel efficient vehicles. Volume I: Executive summary. Final report, October 1977-March 1981

    SciTech Connect

    Heisler, J.T.; Groeneman, S.

    1981-03-01

    The Energy Policy and Conservation Act of 1975 requires fuel efficiency standards through Model Year 1985 and requires that economic practicability (e.g., marketability, consumer impact) be considered in miles per gallon formulations. The purpose of this research is to develop information on likely market response to various vehicle design and performance changes to minimize the possibility of market rejection. The report is based on focus group discussions and consumer experiments with drivers in different vehicle size classes. Analysis focuses on the believability of the 'energy crisis' and concern about its effects, the acceptability of possible fuel economy options, consumer preferences with respect to vehicles embodying selected design and engineering changes, and predicted purchase and usage patterns under different future scenarios.

  6. Zinc air refuelable battery: alternative zinc fuel morphologies and cell behavior

    SciTech Connect

    Cooper, J.F.; Krueger, R.

    1997-01-01

    Multicell zinc/air batteries have been tested previously in the laboratory and as part of the propulsion system of an electric bus; cut zinc wire was used as the anode material. This battery is refueled by a hydraulic transport of 0.5-1 mm zinc particles into hoppers above each cell. We report an investigation concerning alternative zinc fuel morphologies, and energy losses associated with refueling and with overnight or prolonged standby. Three types of fuel pellets were fabricated, tested and compared with results for cut wire: spheres produced in a fluidized bed electrolysis cell; elongated particles produced by gas-atomization; and pellets produced by chopping 1 mm porous plates made of compacted zinc fines. Relative sizes of the particles and cell gap dimensions are critical. All three types transported within the cell 1553 and showed acceptable discharge characteristics, but a fluidized bed approach appears especially attractive for owner/user recovery operations.

  7. 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.

  8. Corrosion behavior of iron and nickel base alloys under solid oxide fuel cell exposure conditions

    SciTech Connect

    Ziomek-Moroz, M.; Holcomb, G.R.; Covino, B.S., Jr.; Bullard, S.J.

    2006-03-01

    Topography and phase composition of the scales formed on commercial ferritic stainless steels and experimental low CTE nickel-based alloys were studied in atmospheres simulating solid oxide fuel cell (SOFC) environments. The materials were studied under dual environment conditions with air on one side of the sample and carbon monoxide on the other side at 750°C. Surface characterization techniques, such as scanning electron microscopy and X-ray diffraction analysis were used in this study.

  9. Phenomenological Study of the Behavior of Some Silica Formers in a High Velocity Jet Fuel Burner.

    DTIC Science & Technology

    1985-10-01

    Svehla, R.A.; and Lewandowski, K.: Thermodynamics and Transport Properties of Air and the Combustion Products of Natural Gas and of ASTM-A-1 Fuel...gas flow through the engine. The materials studied were; molybdenum disilicide , single crystal silicon jL.U carbide, sintered alpha silicon carbide, and...inch) outside of this a steep thermal gradient existed. Sighting of the pyrometer may have contributed to the scat- RESULTS Molybdenum Disilicide Two

  10. Irradiation behavior study of U-Mo/Al dispersion fuel with high energy Xe

    NASA Astrophysics Data System (ADS)

    Ye, B.; Bhattacharya, S.; Mo, K.; Yun, D.; Mohamed, W.; Pellin, M.; Fortner, J.; Kim, Y. S.; Hofman, G. L.; Yacout, A. M.; Wiencek, T.; Van den Berghe, S.; Leenaers, A.

    2015-09-01

    Irradiation responses of U-Mo/Al dispersion fuel have been investigated by irradiation with 84 MeV Xe26+ ions. Dispersion fuels fabricated with uncoated and ZrN-coated fuel particles were irradiated to various doses at ∼350 °C. The highest dose achieved was 2.9 × 1017 ions/cm2 (∼1200 displacement per atom (dpa)). Following the irradiation, scanning electron microscopy (SEM) and transmission electron microscopy (TEM) experiments were carried out to characterize the microstructures of the irradiated samples. The post irradiation examinations (PIE) revealed that: (1) crystalline interdiffusion product (UMo)Alx developed at locations where no coating or compromised coating layer is present; (2) intact ZrN coating layers effectively blocked the interdiffusion between U-Mo and Al; (3) SEM-observable Xe bubbles distributed along grain/cell boundaries in U-Mo; and (4) gas bubble interlinkage was observed at a dose of 2.9 × 1017 ions/cm2.

  11. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

    NASA Astrophysics Data System (ADS)

    Courty, Olivier; Motta, Arthur T.; Hales, Jason D.

    2014-09-01

    As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick's law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

  12. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

    SciTech Connect

    Jason D. Hales; Various

    2014-09-01

    As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick’s law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

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

    SciTech Connect

    Z. Ceylan

    1996-10-18

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

  14. 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.

  15. 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)

  16. FRAPCON-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup

    SciTech Connect

    Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.

    1997-12-01

    FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes` integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6).

  17. Synchronization behavior in linear arrays of negative differential resistance circuit elements

    NASA Astrophysics Data System (ADS)

    Xu, Huidong; Teitsworth, Stephen

    2008-03-01

    We study the electronic transport properties in a linear array of nonlinear circuit elements that exhibit negative differential resistance, and find that state-cluster synchronization emerges when there is heterogeneity in the element properties. This type of synchronization is associated with a non-uniform spatial distribution of total applied voltage across the array elements, as well as the formation of multiple stable branches in computed current-voltage curves for the entire array. Unlike most synchronizing systems studied previously [1], this system possesses coupling between elements that displays both positive and negative feedback depending on the state of each element. An empirical order parameter is defined which quantifies the degree of synchronization. We also find that the degree of synchronization is strongly dependent on the ramping rate of the total applied voltage to the array, with complete synchronization observed in the limit of small ramping rate. This model provides a basis for describing related nonlinear phenomena in more complex electronic structures such as semiconductor superlattices [2]. [1] A. Pikovsky, M. Rosenbaum, and Jürgen Kurths, Synchronization: a universal concept in nonlinear sciences (Cambridge University Press, Cambridge, 2001). [2] M. Rogozia, S. W. Teitsworth, H. T Grahn, and K. H. Ploog, Phys. Rev. B65, 205303 (2002).

  18. 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)

  19. Analysis of the geometrical dependence of auxetic behavior in reentrant structures by finite elements

    NASA Astrophysics Data System (ADS)

    Carneiro, V. H.; Puga, H.; Meireles, J.

    2016-04-01

    Materials with a negative Poisson's ratio (PR) are called auxetics; they are characterized by expansion/contraction when tensioned/compressed. Given this counterintuitive behavior, they present very particular characteristics and mechanical behavior. Geometrical models have been developed to justify and artificially reproduce such materials' auxetic behavior. The focus of this study is the exploration of a reentrant model by analyzing the variation in the PR of reentrant structures as a function of geometrical and base material parameters. It is shown that, even in the presence of protruding ribs, there may not be auxetic behavior, and this depends on the geometry of each reentrant structure. Values determined for these parameters can be helpful as approximate reference data in the design and fabrication of auxetic lattices using reentrant geometries.

  20. 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.