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Sample records for fuel element heat

  1. 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.

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

    NASA Technical Reports Server (NTRS)

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

    2013-01-01

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

  3. FUEL ELEMENT

    DOEpatents

    Bean, R.W.

    1963-11-19

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

  4. 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.

  5. FUEL ELEMENT

    DOEpatents

    Fortescue, P.; Zumwalt, L.R.

    1961-11-28

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

  6. 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.

  7. 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)

  8. 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.

  9. 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)

  10. 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.

  11. 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.

  12. 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.

  13. FUEL ELEMENT SUPPORT

    DOEpatents

    Wyman, W.L.

    1961-06-27

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

  14. 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.

  15. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Dickson, J.J.

    1963-09-24

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

  16. 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)

  17. 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

  18. 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)

  19. 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.

  20. 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.

  1. 3D modeling of heat transfer and gas flow in a grooved ring fuel element for nuclear thermal propulsion

    NASA Astrophysics Data System (ADS)

    Barkett, Laura Ashley

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

  2. 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)

  3. 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.

  4. 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.

  5. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

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

    1958-12-01

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

  6. 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)

  7. 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)

  8. 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)

  9. 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.

  10. 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.

  11. Heating element support clip

    DOEpatents

    Sawyer, W.C.

    1995-08-15

    An apparatus for supporting a heating element in a channel formed in a heater base is disclosed. A preferred embodiment includes a substantially U-shaped tantalum member. The U-shape is characterized by two substantially parallel portions of tantalum that each have an end connected to opposite ends of a base portion of tantalum. The parallel portions are each substantially perpendicular to the base portion and spaced apart a distance not larger than a width of the channel and not smaller than a width of a graphite heating element. The parallel portions each have a hole therein, and the centers of the holes define an axis that is substantially parallel to the base portion. An aluminum oxide ceramic retaining pin extends through the holes in the parallel portions and into a hole in a wall of the channel to retain the U-shaped member in the channel and to support the graphite heating element. The graphite heating element is confined by the parallel portions of tantalum, the base portion of tantalum, and the retaining pin. A tantalum tube surrounds the retaining pin between the parallel portions of tantalum. 6 figs.

  12. Heating element support clip

    DOEpatents

    Sawyer, William C.

    1995-01-01

    An apparatus for supporting a heating element in a channel formed in a heater base is disclosed. A preferred embodiment includes a substantially U-shaped tantalum member. The U-shape is characterized by two substantially parallel portions of tantalum that each have an end connected to opposite ends of a base portion of tantalum. The parallel portions are each substantially perpendicular to the base portion and spaced apart a distance not larger than a width of the channel and not smaller than a width of a graphite heating element. The parallel portions each have a hole therein, and the centers of the holes define an axis that is substantially parallel to the base portion. An aluminum oxide ceramic retaining pin extends through the holes in the parallel portions and into a hole in a wall of the channel to retain the U-shaped member in the channel and to support the graphite heating element. The graphite heating element is confined by the parallel portions of tantalum, the base portion of tantalum, and the retaining pin. A tantalum tube surrounds the retaining pin between the parallel portions of tantalum.

  13. 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.

  14. FUEL ELEMENT FABRICATION METHOD

    DOEpatents

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

    1960-05-31

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

  15. 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)

  16. NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1962-08-14

    A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

  17. Nuclear fuel 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.

  18. 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.

  19. 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.

  20. NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

    DOEpatents

    Szilard, L.; Young, G.J.

    1958-03-01

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

  1. 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)

  2. 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.

  3. FUEL ELEMENTS FOR NUCLEAR REACTORS

    DOEpatents

    Blainey, A.; Lloyd, H.

    1961-07-11

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

  4. 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

  5. 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.

  6. Thermoelectric heat exchange element

    DOEpatents

    Callas, James J.; Taher, Mahmoud A.

    2007-08-14

    A thermoelectric heat exchange module includes a first substrate including a heat receptive side and a heat donative side and a series of undulatory pleats. The module may also include a thermoelectric material layer having a ZT value of 1.0 or more disposed on at least one of the heat receptive side and the heat donative side, and an electrical contact may be in electrical communication with the thermoelectric material layer.

  7. 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.

  8. 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)

  9. 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)

  10. 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)

  11. 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.

  12. 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.

  13. 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.

  14. 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)

  15. 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)

  16. 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.

  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. 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.

  19. 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.

  20. 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.

  1. Self supporting heat transfer element

    DOEpatents

    Story, Grosvenor Cook; Baldonado, Ray Orico

    2002-01-01

    The present invention provides an improved internal heat exchange element arranged so as to traverse the inside diameter of a container vessel such that it makes good mechanical contact with the interior wall of that vessel. The mechanical element is fabricated from a material having a coefficient of thermal conductivity above about 0.8 W cm.sup.-1.degree. K.sup.-1 and is designed to function as a simple spring member when that member has been cooled to reduce its diameter to just below that of a cylindrical container or vessel into which it is placed and then allowed to warm to room temperature. A particularly important application of this invention is directed to a providing a simple compartmented storage container for accommodating a hydrogen absorbing alloy.

  2. 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.

  3. Heating subsurface formations by oxidizing fuel on a fuel carrier

    DOEpatents

    Costello, Michael; Vinegar, Harold J.

    2012-10-02

    A method of heating a portion of a subsurface formation includes drawing fuel on a fuel carrier through an opening formed in the formation. Oxidant is supplied to the fuel at one or more locations in the opening. The fuel is combusted with the oxidant to provide heat to the formation.

  4. 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.

  5. Visual examinations of K east fuel elements

    SciTech Connect

    Pitner, A.L., Fluor Daniel Hanford

    1997-02-03

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

  6. Experiments on fuel heating for commercial aircraft

    NASA Technical Reports Server (NTRS)

    Friedman, R.; Stockemer, F. J.

    1982-01-01

    An experimental jet fuel with a -33 C freezing point was chilled in a wing tank simulator with superimposed fuel heating to improve low temperature flowability. Heating consisted of circulating a portion of the fuel to an external heat exchanger and returning the heated fuel to the tank. Flowability was determined by the mass percent of unpumpable fuel (holdup) left in the simulator upon withdrawal of fuel at the conclusion of testing. The study demonstrated that fuel heating is feasible and improves flowability as compared to that of baseline, unheated tests. Delayed heating with initiation when the fuel reaches a prescribed low temperature limit, showed promise of being more efficient than continuous heating. Regardless of the mode or rate of heating, complete flowability (zero holdup) could not be restored by fuel heating. The severe, extreme-day environment imposed by the test caused a very small amount of subfreezing fuel to be retained near the tank surfaces even at high rates of heating. Correlations of flowability established for unheated fuel tests also could be applied to the heated test results if based on boundary-layer temperature or a solid index (subfreezing point) characteristic of the fuel.

  7. 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)

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. Heat exchanger with ceramic elements

    DOEpatents

    Corey, John A.

    1986-01-01

    An annular heat exchanger assembly includes a plurality of low thermal growth ceramic heat exchange members with inlet and exit flow ports on distinct faces. A mounting member locates each ceramic member in a near-annular array and seals the flow ports on the distinct faces into the separate flow paths of the heat exchanger. The mounting member adjusts for the temperature gradient in the assembly and the different coefficients of thermal expansion of the members of the assembly during all operating temperatures.

  13. Numerical simulation of gas dynamics and heat exchange tasks in fuel assemblies of the nuclear reactors

    SciTech Connect

    Zhuchenko, S. V.

    2014-11-12

    This report presents a PC-based program for solution gas dynamics and heat exchange mathematical tasks in fuel assemblies of the fast-neutron nuclear reactors. A fuel assembly consisting of bulk heat-generating elements, which are integrated together by the system of supply and pressure manifolds, is examined. Spherical heat-generating microelements, which contain nuclear fuel, are pulled into the heat-generating elements. Gaseous coolant proceed from supply manifolds to heat-generating elements, where it withdraws the nuclear reaction heat and assembles in pressure manifolds.

  14. 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.

  15. 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.

  16. Element-by-element factorization algorithms for heat conduction

    NASA Technical Reports Server (NTRS)

    Hughes, T. J. R.; Winget, J. M.; Park, K. C.

    1983-01-01

    Element-by-element solution strategies are developed for transient heat conduction problems. Results of numerical tests indicate the effectiveness of the procedures proposed. The small database requirements and attractive architectural features of the algorithms suggest considerable potential for solving large scale problems.

  17. Volume reduction of spent fuel elements for direct disposal

    SciTech Connect

    Wasserfuhr, I.C.

    1995-12-31

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

  18. 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.

  19. 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.

  20. Local Heat Flux Measurements with Single Element Coaxial Injectors

    NASA Technical Reports Server (NTRS)

    Jones, Gregg; Protz, Christopher; Bullard, Brad; Hulka, James

    2006-01-01

    To support the mission for the NASA Vision for Space Exploration, the NASA Marshall Space Flight Center conducted a program in 2005 to improve the capability to predict local thermal compatibility and heat transfer in liquid propellant rocket engine combustion devices. The ultimate objective was to predict and hence reduce the local peak heat flux due to injector design, resulting in a significant improvement in overall engine reliability and durability. Such analyses are applicable to combustion devices in booster, upper stage, and in-space engines, as well as for small thrusters with few elements in the injector. In this program, single element and three-element injectors were hot-fire tested with liquid oxygen and ambient temperature gaseous hydrogen propellants at The Pennsylvania State University Cryogenic Combustor Laboratory from May to August 2005. Local heat fluxes were measured in a 1-inch internal diameter heat sink combustion chamber using Medtherm coaxial thermocouples and Gardon heat flux gauges. Injectors were tested with shear coaxial and swirl coaxial elements, including recessed, flush and scarfed oxidizer post configurations, and concentric and non-concentric fuel annuli. This paper includes general descriptions of the experimental hardware, instrumentation, and results of the hot-fire testing for three of the single element injectors - recessed-post shear coaxial with concentric fuel, flush-post swirl coaxial with concentric fuel, and scarfed-post swirl coaxial with concentric fuel. Detailed geometry and test results will be published elsewhere to provide well-defined data sets for injector development and model validatation.

  1. 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.

  2. A high-temperature heat sensitive element

    NASA Technical Reports Server (NTRS)

    Oguro, M.

    1986-01-01

    This invention concerns the high-temperature heat sensitive element which is stable at high temperatures. A solid solution of the main component MgO-Al2O3-Cr2O3-Fe2O3 which contains spinel crystal structure is mixed with the secondary component ZrO2 at the mol ratio of 100 : 0.1 to 5.0 and sintered to prepare a high-temperature heat sensitive element.

  3. 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.

  4. Identification of leaking TRIGA fuel elements

    SciTech Connect

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

    1990-07-01

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

  5. 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.

  6. Design of heat exchange element for plastic film heat exchanger

    NASA Astrophysics Data System (ADS)

    Guyer, E. C.; Brownell, D. L.

    1984-12-01

    This report presents the results of an effort to design a plastic film heat exchanger element (PFHX) suitable for use in an industrial heat pump evaporator. This report addresses the selection of materials, the expected flow and heat transfer behavior, and the mechanical design features of a parallel plate type exchanger that uses thin plastic films as the boundary between the two process fluids. Criteria for material selection are presented, candidate materials are reviewed, and material recommendations are provided. Heat transfer performance is addressed in terms of the overall or total coefficient of heat transfer between condensing steam and a confined falling film of water. Appropriate mechanical designs of water flow manifolds are described along with methods of fabrication and assembly. This report addresses only the individual heat exchange element.

  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. REGENERATION OF REACTOR FUEL ELEMENTS

    DOEpatents

    Lyon, W.L.

    1960-04-01

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

  9. 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.

  10. Upgraded HFIR Fuel Element Welding System

    SciTech Connect

    Sease, John D

    2010-02-01

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

  11. Nuclear fuel elements made from nanophase materials

    SciTech Connect

    Heubeck, Norman B.

    1997-12-01

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain-related failure even at high temperatures, in the order of about 3,000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion and mechanical characteristics.

  12. Nuclear fuel elements made from nanophase materials

    DOEpatents

    Heubeck, Norman B.

    1998-01-01

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

  13. 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.

  14. 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.

  15. 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.

  16. Fuel delivery system including heat exchanger means

    NASA Technical Reports Server (NTRS)

    Coffinberry, G. A. (Inventor)

    1978-01-01

    A fuel delivery system is presented wherein first and second heat exchanger means are each adapted to provide the transfer of heat between the fuel and a second fluid such as lubricating oil associated with the gas turbine engine. Valve means are included which are operative in a first mode to provide for flow of the second fluid through both first and second heat exchange means and further operative in a second mode for bypassing the second fluid around the second heat exchanger means.

  17. Heating Values of Fuels: An Introductory Experiment.

    ERIC Educational Resources Information Center

    Rettlich, Timothy R.; And Others

    1988-01-01

    Describes a simple, inexpensive experiment in which students determine the heats of combustion of common solid, liquid, and gaseous fuels. The experimental apparatus, procedures, calculations and results are discussed. (CW)

  18. Heat Exchanger With Internal Pin Elements

    DOEpatents

    Gerstmann, Joseph; Hannon, Charles L.

    2004-01-13

    A heat exchanger/heater comprising a tubular member having a fluid inlet end, a fluid outlet end and plurality of pins secured to the interior wall of the tube. Various embodiments additionally comprise a blocking member disposed concentrically inside the pins, such as a core plug or a baffle array. Also disclosed is a vapor generator employing an internally pinned tube, and a fluid-heater/heat-exchanger utilizing an outer jacket tube and fluid-side baffle elements, as well as methods for heating a fluid using an internally pinned tube.

  19. Code System for Spent Fuel Heating Analysis.

    Energy Science and Technology Software Center (ESTSC)

    1999-05-24

    Version 00 SFHA calculates steady-state fuel rod temperatures for hexagon and square-fuel bundles. The code is used to perform sensitivity studies and confirmatory analyses of results submitted by applicants for spent fuel storage licenses. All three modes of heat transfer are considered; radiation, convection, and conduction. Each is modeled separately. SFHA benchmark calculations were made with test data to validate the use of a simple one-dimensional heat transfer model for estimating fuel rod temperatures. Benchmarkmore » results show that SFHA is capable of calculating spent fuel rod temperatures for square and hexagonal fuel bundles under various environments for the consolidated or unconsolidated condition. The program is menu-driven and executes automatically after all required information is entered.« less

  20. Heat Transfer to Fuel Sprays Injected into Heated Gases

    NASA Technical Reports Server (NTRS)

    Selden, Robert F; Spencer, Robert C

    1938-01-01

    This report presents the results of a study made of the influence of several variables on the pressure decrease accompanying injection of a relatively cool liquid into a heated compressed gas. Indirectly, this pressure decrease and the time rate of change of it are indicative of the total heat transferred as well as the rate of heat transfer between the gas and the injected liquid. Air, nitrogen, and carbon dioxide were used as ambient gases; diesel fuel and benzene were the injected liquids. The gas densities and gas-fuel ratios covered approximately the range used in compression-ignition engines. The gas temperatures ranged from 150 degrees c. to 350 degrees c.

  1. Interface elements for heat transfer analysis

    NASA Astrophysics Data System (ADS)

    Mason, W. E.

    1984-08-01

    Interface elements are desirable in finite element heat transfer analyses in situations where dissimilar meshes are to be joined or where contact resistances occur between various parts of a body. In stress codes, such elements are often termed master/slave. A general algorithm for interface elements will be described. The algorithm allows development of interface elements for both two- and three-dimensional applications. Surfaces in contact are automatically determined so that a minimum of input data is required. In addition, the algorithm allows for compatibility in thermal stress calculations with mechanical codes which have sliding interface capabilities. Implementation of the algorithm into the TACO codes will be discussed and examples will be given.

  2. Liquid fuel injection elements for rocket engines

    NASA Technical Reports Server (NTRS)

    Cox, George B., Jr. (Inventor)

    1993-01-01

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

  3. Automatic inspection for remotely manufactured fuel elements

    SciTech Connect

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

    1995-06-01

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

  4. Thermionic Fuel Element Verification Program - Overview

    NASA Astrophysics Data System (ADS)

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

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

  5. Coupled thermionic and thermalhydraulic analyses of thermionic fuel elements

    NASA Astrophysics Data System (ADS)

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

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

  6. Finite Element Heat & Mass Transfer Code

    Energy Science and Technology Software Center (ESTSC)

    1996-10-10

    FEHM is a numerical simulation code for subsurface transport processes. It models 3-D, time-dependent, multiphase, multicomponent, non-isothermal, reactive flow through porous and fractured media. It can accurately represent complex 3-D geologic media and structures and their effects on subsurface flow and transport. Its capabilities include flow of gas, water, and heat; flow of air, water, and heat; multiple chemically reactive and sorbing tracers; finite element/finite volume formulation; coupled stress module; saturated and unsaturated media; andmore » double porosity and double porosity/double permeability capabilities.« less

  7. METHOD OF MAKING WIRE FUEL ELEMENTS

    DOEpatents

    Zambrow, J.L.

    1960-08-01

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

  8. 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)

  9. Shielded regeneration heating element for a particulate filter

    DOEpatents

    Gonze, Eugene V [Pinckney, MI; Ament, Frank [Troy, MI

    2011-01-04

    An exhaust system includes a particulate filter (PF) that is disposed downstream from an engine. The PF filters particulates within an exhaust from the engine. A heating element heats particulate matter in the PF. A catalyst substrate or a flow converter is disposed upstream from said heating element. The catalyst substrate oxidizes the exhaust prior to reception by the heating element. The flow converter converts turbulent exhaust flow to laminar exhaust flow prior to reception by the heating element.

  10. ITER Ion Cyclotron Heating and Fueling Systems

    SciTech Connect

    Rasmussen, D.A.; Baylor, L.R.; Combs, S.K.; Fredd, E.; Goulding, R.H.; Hosea, J.; Swain, D.W.

    2005-04-15

    The ITER burning plasma and advanced operating regimes require robust and reliable heating and current drive and fueling systems. The ITER design documents describe the requirements and reference designs for the ion cyclotron and pellet fueling systems. Development and testing programs are required to optimize, validate and qualify these systems for installation on ITER.The ITER ion cyclotron system offers significant technology challenges. The antenna must operate in a nuclear environment and withstand heat loads and disruption forces beyond present-day designs. It must operate for long pulse lengths and be highly reliable, delivering power to a plasma load with properties that will change throughout the discharge. The ITER ion cyclotron system consists of one eight-strap antenna, eight rf sources (20 MW, 35-65 MHz), associated high-voltage DC power supplies, transmission lines and matching and decoupling components.The ITER fueling system consists of a gas injection system and multiple pellet injectors for edge fueling and deep core fueling. Pellet injection will be the primary ITER fuel delivery system. The fueling requirements will require significant extensions in pellet injector pulse length ({approx}3000 s), throughput (400 torr-L/s,) and reliability. The proposed design is based on a centrifuge accelerator fed by a continuous screw extruder. Inner wall pellet injection with the use of curved guide tubes will be utilized for deep fueling.

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

    NASA Astrophysics Data System (ADS)

    Shaheen, Khaled

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

  12. 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)

  13. Heat exchanger for fuel cell power plant reformer

    DOEpatents

    Misage, Robert; Scheffler, Glenn W.; Setzer, Herbert J.; Margiott, Paul R.; Parenti, Jr., Edmund K.

    1988-01-01

    A heat exchanger uses the heat from processed fuel gas from a reformer for a fuel cell to superheat steam, to preheat raw fuel prior to entering the reformer and to heat a water-steam coolant mixture from the fuel cells. The processed fuel gas temperature is thus lowered to a level useful in the fuel cell reaction. The four temperature adjustments are accomplished in a single heat exchanger with only three heat transfer cores. The heat exchanger is preheated by circulating coolant and purge steam from the power section during startup of the latter.

  14. 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)

  15. Thermionic fuel element Verification Program - Overview

    NASA Astrophysics Data System (ADS)

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

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

  16. Thermionic fuel element verification program—overview

    NASA Astrophysics Data System (ADS)

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

    1991-01-01

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

  17. 3-D Finite Element Heat Transfer

    Energy Science and Technology Software Center (ESTSC)

    1992-02-01

    TOPAZ3D is a three-dimensional implicit finite element computer code for heat transfer analysis. TOPAZ3D can be used to solve for the steady-state or transient temperature field on three-dimensional geometries. Material properties may be temperature-dependent and either isotropic or orthotropic. A variety of time-dependent and temperature-dependent boundary conditions can be specified including temperature, flux, convection, and radiation. By implementing the user subroutine feature, users can model chemical reaction kinetics and allow for any type of functionalmore » representation of boundary conditions and internal heat generation. TOPAZ3D can solve problems of diffuse and specular band radiation in an enclosure coupled with conduction in the material surrounding the enclosure. Additional features include thermal contact resistance across an interface, bulk fluids, phase change, and energy balances.« less

  18. Fuel cell system with combustor-heated reformer

    DOEpatents

    Pettit, William Henry

    2000-01-01

    A fuel cell system including a fuel reformer heated by a catalytic combustor fired by anode effluent and/or fuel from a liquid fuel supply providing fuel for the fuel cell. The combustor includes a vaporizer section heated by the combustor exhaust gases for vaporizing the fuel before feeding it into the combustor. Cathode effluent is used as the principle oxidant for the combustor.

  19. 2-D Finite Element Heat Conduction

    Energy Science and Technology Software Center (ESTSC)

    1989-10-30

    AYER is a finite element program which implicitly solves the general two-dimensional equation of thermal conduction for plane or axisymmetric bodies. AYER takes into account the effects of time (transient problems), in-plane anisotropic thermal conductivity, a three-dimensional velocity distribution, and interface thermal contact resistance. Geometry and material distributions are arbitrary, and input is via subroutines provided by the user. As a result, boundary conditions, material properties, velocity distributions, and internal power generation may be mademore » functions of, e.g., time, temperature, location, and heat flux.« less

  20. Microfabricated fuel heating value monitoring device

    DOEpatents

    Robinson, Alex L.; Manginell, Ronald P.; Moorman, Matthew W.

    2010-05-04

    A microfabricated fuel heating value monitoring device comprises a microfabricated gas chromatography column in combination with a catalytic microcalorimeter. The microcalorimeter can comprise a reference thermal conductivity sensor to provide diagnostics and surety. Using microfabrication techniques, the device can be manufactured in production quantities at a low per-unit cost. The microfabricated fuel heating value monitoring device enables continuous calorimetric determination of the heating value of natural gas with a 1 minute analysis time and 1.5 minute cycle time using air as a carrier gas. This device has applications in remote natural gas mining stations, pipeline switching and metering stations, turbine generators, and other industrial user sites. For gas pipelines, the device can improve gas quality during transfer and blending, and provide accurate financial accounting. For industrial end users, the device can provide continuous feedback of physical gas properties to improve combustion efficiency during use.

  1. Method and apparatus for fuel gas moisturization and heating

    DOEpatents

    Ranasinghe, Jatila; Smith, Raub Warfield

    2002-01-01

    Fuel gas is saturated with water heated with a heat recovery steam generator heat source. The heat source is preferably a water heating section downstream of the lower pressure evaporator to provide better temperature matching between the hot and cold heat exchange streams in that portion of the heat recovery steam generator. The increased gas mass flow due to the addition of moisture results in increased power output from the gas and steam turbines. Fuel gas saturation is followed by superheating the fuel, preferably with bottom cycle heat sources, resulting in a larger thermal efficiency gain compared to current fuel heating methods. There is a gain in power output compared to no fuel heating, even when heating the fuel to above the LP steam temperature.

  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. METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

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

    1960-06-21

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

  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. 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)

  6. Study of fuel cell powerplant with heat recovery

    NASA Technical Reports Server (NTRS)

    King, J. M.; Grasso, A. P.; Clausi, J. V.

    1975-01-01

    It was shown that heat can be recovered from fuel cell power plants by replacing the air-cooled heat exchangers in present designs with units which transfer the heat to the integrated utility system. Energy availability for a 40-kW power plant was studied and showed that the total usable energy at rated power represents 84 percent of the fuel lower heating value. The effects of design variables on heat availability proved to be small. Design requirements were established for the heat recovery heat exchangers, including measurement of the characteristics of two candidate fuel cell coolants after exposure to fuel cell operating conditions. A heat exchanger test program was defined to assess fouling and other characteristics of fuel cell heat exchangers needed to confirm heat exchanger designs for heat recovery.

  7. 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.

  8. TACO: a finite element heat transfer code

    SciTech Connect

    Mason, W.E. Jr.

    1980-02-01

    TACO is a two-dimensional implicit finite element code for heat transfer analysis. It can perform both linear and nonlinear analyses and can be used to solve either transient or steady state problems. Either plane or axisymmetric geometries can be analyzed. TACO has the capability to handle time or temperature dependent material properties and materials may be either isotropic or orthotropic. A variety of time and temperature dependent loadings and boundary conditions are available including temperature, flux, convection, and radiation boundary conditions and internal heat generation. Additionally, TACO has some specialized features such as internal surface conditions (e.g., contact resistance), bulk nodes, enclosure radiation with view factor calculations, and chemical reactive kinetics. A user subprogram feature allows for any type of functional representation of any independent variable. A bandwidth and profile minimization option is also available in the code. Graphical representation of data generated by TACO is provided by a companion post-processor named POSTACO. The theory on which TACO is based is outlined, the capabilities of the code are explained, the input data required to perform an analysis with TACO are described. Some simple examples are provided to illustrate the use of the code.

  9. Local Measurement of Fuel Energy Deposition and Heat Transfer Environment During Fuel Lifetime Using Controlled Calorimetry

    SciTech Connect

    Don W. Miller; Andrew Kauffmann; Eric Kreidler; Dongxu Li; Hanying Liu; Daniel Mills; Thomas D. Radcliff; Joseph Talnagi

    2001-12-31

    A comprehensive description of the accomplishments of the DOE grant titled, ''Local Measurement of Fuel Energy Deposition and Heat Transfer Environment During Fuel Lifetime using Controlled Calorimetry''.

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

    Energy Science and Technology Software Center (ESTSC)

    2001-09-24

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

  11. 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.

  12. 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.

  13. 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

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

    NASA Technical Reports Server (NTRS)

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

    2013-01-01

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

  15. 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.

  16. 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.

  17. Reaction heat used in static water removal from fuel cells

    NASA Technical Reports Server (NTRS)

    Platner, J. L.

    1966-01-01

    Reaction heat is used for removal of water formed at the hydrogen fuel electrode in a hydrogen-oxygen fuel cell. A portion of the heat inherent in the fuel cell current generation reaction is used to transfer excess water into water vapor and cause it to be exhausted from the cell by a porous vapor transport membrane adjoining a vapor cavity.

  18. Fuel type impact at heat exchanger performance

    NASA Astrophysics Data System (ADS)

    Durčanský, Peter; Patsch, Marek; Jandačka, Jozef

    2016-06-01

    Possible solution to the increasing energy consumption in the world may be use of alternative energy sources in micro-cogeneration in combination with increasing energy effectiveness. The use of renewable sources, such as biomass, represents an important contribution to possible solution of this problem. When designing a new heat source it is required to follow a number of technical regulations and recommendations. The proposed combustion furnace is intended for combustion of biomass, either piece, or in the form of wood biomass. But the combustion is not only affected by design of furnace, but also by fuel and its properties.

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

    SciTech Connect

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

    1997-03-06

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

  20. 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/.

  1. Heat transfer to a supercritical hydrocarbon fuel with endothermic reaction.

    SciTech Connect

    Yu, W.; France, D. M.; Wambsganss, M. W.; Energy Technology; Univ. of Illinois at Chicago

    2000-01-01

    Supercritical fuel reforming is being studied as a technology for reducing emissions of industrial gas turbine engines. In this study, experiments were performed in a 2.67-mm-inside-diameter stainless steel tube with a heated length of 0.610 m for the purpose of investigating the characteristics of supercritical heat transfer with endothermic fuel reforming. Thermocouples were positioned along the tube both in the fluid stream and on the heated wall for local heat transfer measurements. Both heat transfer coefficients and endotherms were calculated from the measured results. State-of-the-art correlations for heat transfer were evaluated, and a correlation for supercritical heat transfer to hydrocarbon fuel has been developed. The results provide a basis for supercritical fuel heat-exchanger/reactor design and its practical applications, in an area that has received relatively little attention in the engineering literature, viz., supercritical forced convection heat transfer with endothermic chemical reaction.

  2. Heat Transfer Variation on Protuberances and Surface Roughness Elements

    NASA Technical Reports Server (NTRS)

    Henry, Robert C.; Hansman, R. John, Jr.; Breuer, Kenneth S.

    1995-01-01

    In order to determine the effect of surface irregularities on local convective heat transfer, the variation in heat transfer coefficients on small (2-6 mm diam) hemispherical roughness elements on a flat plate has been studied in a wind funnel using IR techniques. Heat transfer enhancement was observed to vary over the roughness elements with the maximum heat transfer on the upstream face. This heat transfer enhancement increased strongly with roughness size and velocity when there was a laminar boundary layer on the plate. For a turbulent boundary layer, the heat transfer enhancement was relatively constant with velocity, but did increase with element size. When multiple roughness elements were studied, no influence of adjacent roughness elements on heat transfer was observed if the roughness separation was greater than approximately one roughness element radius. As roughness separation was reduced, less variation in heat transfer was observed on the downstream elements. Implications of the observed roughness enhanced heat transfer on ice accretion modeling are discussed.

  3. VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

    DOEpatents

    Furgerson, W.T.

    1963-12-17

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

  4. Combined Heat and Power Market Potential for Opportunity Fuels

    SciTech Connect

    Jones, David; Lemar, Paul

    2015-12-01

    This report estimates the potential for opportunity fuel combined heat and power (CHP) applications in the United States, and provides estimates for the technical and economic market potential compared to those included in an earlier report. An opportunity fuel is any type of fuel that is not widely used when compared to traditional fossil fuels. Opportunity fuels primarily consist of biomass fuels, industrial waste products and fossil fuel derivatives. These fuels have the potential to be an economically viable source of power generation in various CHP applications.

  5. 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.

  6. Moving-Temperature-Gradient Heat-Pipe Furnace Element

    NASA Technical Reports Server (NTRS)

    Gillies, Donald C.; Lehoczky, Sandor L.; Gernert, Nelson J.

    1993-01-01

    In improved apparatus, ampoule of material directionally solidified mounted in central hole of annular heat pipe, at suitable axial position between heated and cooled ends. Heated end held in fixed position in single-element furnace; other end left in ambient air or else actively cooled. Gradient of temperature made to move along heat pipe by changing pressure of noncondensable gas. In comparison with prior crystal-growing apparatuses, this one simpler, smaller, and more efficient.

  7. 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.

  8. Liquid Metal Bond for Improved Heat Transfer in LWR Fuel Rods

    SciTech Connect

    Donald Olander

    2005-08-24

    A liquid metal (LM) consisting of 1/3 weight fraction each of Pb, Sn, and Bi has been proposed as the bonding substance in the pellet-cladding gap in place of He. The LM bond eliminates the large AT over the pre-closure gap which is characteristic of helium-bonded fuel elements. Because the LM does not wet either UO2 or Zircaloy, simply loading fuel pellets into a cladding tube containing LM at atmospheric pressure leaves unfilled regions (voids) in the bond. The HEATING 7.3 heat transfer code indicates that these void spaces lead to local fuel hot spots.

  9. 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.

  10. 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.

  11. 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

  12. Control apparatus and method for efficiently heating a fuel processor in a fuel cell system

    DOEpatents

    Doan, Tien M.; Clingerman, Bruce J.

    2003-08-05

    A control apparatus and method for efficiently controlling the amount of heat generated by a fuel cell processor in a fuel cell system by determining a temperature error between actual and desired fuel processor temperatures. The temperature error is converted to a combustor fuel injector command signal or a heat dump valve position command signal depending upon the type of temperature error. Logic controls are responsive to the combustor fuel injector command signals and the heat dump valve position command signal to prevent the combustor fuel injector command signal from being generated if the heat dump valve is opened or, alternately, from preventing the heat dump valve position command signal from being generated if the combustor fuel injector is opened.

  13. 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.

  14. 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)

  15. Low NOx heavy fuel combustor concept program addendum: Low/mid heating value gaseous fuel evaluation

    NASA Technical Reports Server (NTRS)

    Novick, A. S.; Troth, D. L.

    1982-01-01

    The combustion performance of a rich/quench/lean (RQL) combustor was evaluated when operated on low and mid heating value gaseous fuels. Two synthesized fuels were prepared having lower heating values of 10.2 MJ/cu m. (274 Btu/scf) and 6.6 MJ/cu m (176 Btu/scf). These fuels were configured to be representative of actual fuels, being composed primarily of nitrogen, hydrogen, carbon monoxide, and carbon dioxide. A liquid fuel air assist fuel nozzle was modified to inject both of the gaseous fuels. The RQL combustor liner was not changed from the configuration used when the liquid fuels were tested. Both gaseous fuels were tested over a range of power levels from 50 percent load to maximum rated power of the DDN Model 570-K industrial gas turbine engine. Exhaust emissions were recorded for four power level at several rich zone equivalence ratios to determine NOx sensitivity to the rich zone operating point. For the mid Btu heating value gas, ammonia was added to the fuel to simulate a fuel bound nitrogen type gaseous fuel. Results at the testing showed that for the low heating value fuel NOx emissions were all below 20 ppmc and smoke was below a 10 smoke number. For the mid heating value fuel, NOx emissions were in the 50 to 70 ppmc range with the smoke below a 10 smoke number.

  16. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

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

    1993-01-14

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

  17. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

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

    1993-01-15

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

  18. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

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

    1993-05-01

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

  19. HEAT PUMPS: SUBSTITUTES FOR OUTMODED FOSSIL-FUELED SYSTEMS

    EPA Science Inventory

    The report reviews the state-of-the-art relative to development, capacity, and adequacy of the heat pump as a potential replacement for outmoded fossil-fueled heating and cooling systems in the residential and commercial sector. Projections are made of the rate at which heat pump...

  20. High temperature electrically conducting ceramic heating element and control system

    NASA Technical Reports Server (NTRS)

    Halbach, C. R.; Page, R. J.

    1975-01-01

    Improvements were made in both electrode technology and ceramic conductor quality to increase significantly the lifetime and thermal cycling capability of electrically conducting ceramic heater elements. These elements were operated in vacuum, inert and reducing environments as well as oxidizing atmospheres adding to the versatility of the conducting ceramic as an ohmic heater. Using stabilized zirconia conducting ceramic heater elements, a furnace was fabricated and demonstrated to have excellent thermal response and cycling capability. The furnace was used to melt platinum-20% rhodium alloy (melting point 1904 C) with an isothermal ceramic heating element having a nominal working cavity size of 2.5 cm diameter by 10.0 cm long. The furnace was operated to 1940 C with the isothermal ceramic heating element. The same furnace structure was fitted with a pair of main heater elements to provide axial gradient temperature control over a working cavity length of 17.8 cm.

  1. 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.

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

    SciTech Connect

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

    2013-01-31

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

  3. A fuel-oil matrix heat exchanger

    NASA Astrophysics Data System (ADS)

    Mikulin, E. I.; Shevich, Iu. A.; Potapov, V. N.; Veselov, V. A.; Saltais, E. A.; Glukhovskii, G. I.

    A novel design of a welded matrix heat exchanger capable of handling high-pressure liquid and gas coolants is described. Results of tests conducted on matrix heat exchangers and their models are presented, and formulas are recommended for calculating the heat transfer and hydraulic resistance characteristics. A comparison of the characteristics of matrix and tube heat exchangers demonstrates the advantages of the former.

  4. Experimental Investigations of Heat and Mass Transfer in Microchannel Heat-Transfer Elements

    NASA Astrophysics Data System (ADS)

    Konovalov, D. A.

    2016-06-01

    The present work seeks to develop and investigate experimentally microchannel heat-exchange apparatuses of two designs: with porous elements manufactured from titanium and copper, and also based on the matrix of filamentary silicon single crystals under operating conditions with high heat loads, unsteadiness, and nonlinear flow of the coolant. For experimental investigations, the authors have developed and manufactured a unique test bench allowing tests of the developed heat-transfer elements in unsteady operating regimes. The performed experimental investigations have made it possible to obtain criterial dependences of the heat-transfer coefficient on the Reynolds and Prandtl numbers and to refine the values of viscous and inertial coefficients. It has been established that microchannel heat-transfer elements based on silicon single crystals, which make it possible to remove a heat flux above 100 W/cm2, are the most efficient. For porous heat-transfer elements, the best result was attained for wedge-shaped copper samples. According to investigation results, the authors have considered the issues of optimization of thermal and hydraulic characteristics of the heat-transfer elements under study. In the work, the authors have given examples of practical use of the developed heat-transfer elements for cooling systems of radioelectronic equipment.

  5. Experimental Investigations of Heat and Mass Transfer in Microchannel Heat-Transfer Elements

    NASA Astrophysics Data System (ADS)

    Konovalov, D. A.

    2016-05-01

    The present work seeks to develop and investigate experimentally microchannel heat-exchange apparatuses of two designs: with porous elements manufactured from titanium and copper, and also based on the matrix of filamentary silicon single crystals under operating conditions with high heat loads, unsteadiness, and nonlinear flow of the coolant. For experimental investigations, the authors have developed and manufactured a unique test bench allowing tests of the developed heat-transfer elements in unsteady operating regimes. The performed experimental investigations have made it possible to obtain criterial dependences of the heat-transfer coefficient on the Reynolds and Prandtl numbers and to refine the values of viscous and inertial coefficients. It has been established that microchannel heat-transfer elements based on silicon single crystals, which make it possible to remove a heat flux above 100 W/cm2, are the most efficient. For porous heat-transfer elements, the best result was attained for wedge-shaped copper samples. According to investigation results, the authors have considered the issues of optimization of thermal and hydraulic characteristics of the heat-transfer elements under study. In the work, the authors have given examples of practical use of the developed heat-transfer elements for cooling systems of radioelectronic equipment.

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

    NASA Astrophysics Data System (ADS)

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

    2008-12-01

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

  7. Fuel Element for a Nuclear Reactor

    DOEpatents

    Duffy, Jr., J. G.

    1961-05-30

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

  8. FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Duffy, J.G. Jr.

    1961-05-30

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

  9. Local Heat Flux Measurements with Single and Small Multi-element Coaxial Element-Injectors

    NASA Technical Reports Server (NTRS)

    Jones, Gregg; Protz, Christopher; Bullard, Brad; Hulka, James

    2006-01-01

    To support NASA's Vision for Space Exploration mission, the NASA Marshall Space Flight Center conducted a program in 2005 to improve the capability to predict local thermal compatibility and heat transfer in liquid propellant rocket engine combustion devices. The ultimate objective was to predict and hence reduce the local peak heat flux due to injector design, resulting in a significant improvement in overall engine reliability and durability. Such analyses are applicable to combustion devices in booster, upper stage, and in-space engines with regeneratively cooled chamber walls, as well as in small thrust chambers with few elements in the injector. In this program, single and three-element injectors were hot-fire tested with liquid oxygen and gaseous hydrogen propellants at The Pennsylvania State University Cryogenic Combustor Laboratory from May to August 2005. Local heat fluxes were measured in a 1-inch internal diameter heat sink combustion chamber using Medtherm coaxial thermocouples and Gardon heat flux gauges, Injector configurations were tested with both shear coaxial elements and swirl coaxial elements. Both a straight and a scarfed single element swirl injector were tested. This paper includes general descriptions of the experimental hardware, instrumentation, and results of the hot-fire testing for three coaxial shear and swirl elements. Detailed geometry and test results the for shear coax elements has already been published. Detailed test result for the remaining 6 swirl coax element for the will be published in a future JANNAF presentation to provide well-defined data sets for development and model validation.

  10. 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.

  11. 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.

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

    SciTech Connect

    LAWRENCE, L.A.

    1998-10-13

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

  13. 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}.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Finite element analysis of advanced neutron source fuel plates

    SciTech Connect

    Luttrell, C.R.

    1995-08-01

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

  19. 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.

  20. Fuel expansion system with preheater and EMI-heated fuel injector

    SciTech Connect

    Goldsberry, J.

    1989-09-05

    This patent describes a fuel expansion, pre-combustion treatment device for use with carburated or fuel injected combustion engines. It comprises sonic heating and fuel line cleansing means; foraminous dispersing means; EMI field generating means for essentially irradiating the dispersing means.

  1. Role of fuel chemical properties on combustor radiative heat load

    NASA Technical Reports Server (NTRS)

    Rosfjord, T. J.

    1984-01-01

    In an attempt to rigorously study the fuel chemical property influence on combustor radiative heat load, United Technologies Research Center (UTRC) has conducted an experimental program using 25 test fuels. The burner was a 12.7-cm dia cylindrical device fueled by a single pressure-atomizing injector. Fuel physical properties were de-emphasized by selecting injectors which produced high-atomized, and hence rapidly-vaporizing sprays. The fuels were specified to cover the following wide ranges of chemical properties; hydrogen, 9.1 to 15- (wt) pct; total aromatics, 0 to 100 (vol) pct; and naphthalene, 0 to 30 (vol) pct. They included standard fuels, specialty products and fuel blends. Fuel naphthalene content exhibited the strongest influence on radiation of the chemical properties investigated. Smoke point was a good global indicator of radiation severity.

  2. Role of fuel chemical properties on combustor radiative heat load

    NASA Technical Reports Server (NTRS)

    Rosfjord, T. J.

    1984-01-01

    In an attempt to rigorously study the fuel chemical property influence on combustor radiative heat load, UTRC has conducted an experimental program using 25 test fuels. The burner was a 12.7-cm dia cylindrical device fueled by a single pressure-atomizing injector. Fuel physical properties were de-emphasized by selecting injectors which produced highly-atomized, and hence rapidly-vaporizing sprays. The fuels were specified to cover the following wide ranges of chemical properties: hydrogen, 9.1 to 15- (wt) pct; total aromatics, 0 to 100 (vol) pct; and naphthalene, 0 to 30 (vol) pct. They included standard fuels, specialty products and fuel blends. Fuel naphthalene content exhibited the strongest influence on radiation of the chemical properties investigated. Smoke point was a good global indicator of radiation severity.

  3. 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)

  4. Electrical and Joule heating relationship investigation using Finite Element Method

    NASA Astrophysics Data System (ADS)

    Thangaraju, S. K.; Munisamy, K. M.

    2015-09-01

    The finite element method is vastly used in material strength analysis. The nature of the finite element solver, which solves the Fourier equation of stress and strain analysis, made it possible to apply for conduction heat transfer Fourier Equation. Similarly the Current and voltage equation is also liner Fourier equation. The nature of the governing equation makes it possible to numerical investigate the electrical joule heating phenomena in electronic component. This paper highlights the Finite Element Method (FEM) application onto semiconductor interconnects to determine the specific contact resistance (SCR). Metal and semiconductor interconnects is used as model. The result confirms the possibility and validity of FEM utilization to investigate the Joule heating due electrical resistance.

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

    NASA Astrophysics Data System (ADS)

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

    2002-03-01

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

  6. Deposit formation and heat transfer in hydrocarbon rocket fuels

    NASA Technical Reports Server (NTRS)

    Giovanetti, A. J.; Spadaccini, L. J.; Szetela, E. J.

    1983-01-01

    An experimental research program was undertaken to investigate the thermal stability and heat transfer characteristics of several hydrocarbon fuels under conditions that simulate high-pressure, rocket engine cooling systems. The rates of carbon deposition in heated copper and nickel-plated copper tubes were determined for RP-1, propane, and natural gas using a continuous flow test apparatus which permitted independent variation and evaluation of the effect on deposit formation of wall temperature, fuel pressure, and fuel velocity. In addition, the effects of fuel additives and contaminants, cryogenic fuel temperatures, and extended duration testing with intermittent operation were examined. Parametric tests to map the thermal stability characteristics of RP-1, commercial-grade propane, and natural gas were conducted at pressures of 6.9 to 13.8 MPa, bulk fuel velocities of 30 to 90 m/s, and tube wall temperatures in the range of 230 to 810 K. Also, tests were run in which propane and natural gas fuels were chilled to 230 and 160 K, respectively. Corrosion of the copper tube surface was detected for all fuels tested. Plating the inside of the copper tubes with nickel reduced deposit formation and eliminated tube corrosion in most cases. The lowest rates of carbon deposition were obtained for natural gas, and the highest rates were obtained for propane. For all fuels tested, the forced-convection heat transfer film coefficients were satisfactorily correlated using a Nusselt-Reynolds-Prandtl number equation.

  7. Measuring Elemental Abundances in Impulsive Heating Events with EIS

    NASA Astrophysics Data System (ADS)

    Warren, Harry; Doschek, George A.; Young, Peter

    2015-04-01

    It is well established that elemental abundances vary in the solar atmosphere and that this variation is organized by first ionization potential (FIP). Previous studies have indicated that in the solar corona low FIP elements, such as Fe, Si, and Mg, are enriched relative to high FIP elements, such as H, He, C, N, and O. In this paper we report on measurements of plasma composition made during transient heating events observed at transition region temperatures with the Extreme Ultraviolet Imaging Spectrometer (EIS) on Hinode. During these events the intensities of O IV, V, and VI emission lines are enhanced relative to emission lines from Mg V, VI, and VII and indicate a composition close to that of the photosphere. Differential emission measure calculations show a broad distribution of temperatures in these events. Long-lived coronal structures, in contrast, show an enrichment of low FIP elements and relatively narrow temperature distributions. We conjecture that plasma composition is an important signature of the coronal heating process, with impulsive heating leading to the evaporation of unfractionated material from the lower layers of the solar atmosphere and higher frequency heating leading to the accumulation of low-FIP elements in the corona.

  8. 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)

  9. 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)

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

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

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

  11. 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

  12. Particulate Emissions Hazards Associated with Fueling Heat Engines

    NASA Technical Reports Server (NTRS)

    Hendricks, Robert C.; Bushnell, Dennis M.

    2010-01-01

    All hydrocarbon- (HC-) fueled heat engine exhaust (tailpipe) emissions (<10 to 140 nm) contribute as health hazards, including emissions from transportation vehicles (e.g., aircraft) and other HC-fueled power systems. CO2 emissions are tracked, and when mapped, show outlines of major transportation routes and cities. Particulate pollution affects living tissue and is found to be detrimental to cardiovascular and respiratory systems where ultrafine particulates directly translocate to promote vascular system diseases potentially detectable as organic vapors. This paper discusses aviation emissions, fueling, and certification issues, including heat engine emissions hazards, detection at low levels and tracking of emissions, and alternate energy sources for general aviation.

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

    SciTech Connect

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

    1994-10-01

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

  14. 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.

  15. VIEW OF THE HEATING ELEMENTS AND VACUUM GAUGE OF A ...

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

    VIEW OF THE HEATING ELEMENTS AND VACUUM GAUGE OF A PUMP-DOWN STATION IN BUILDING 991. THE PUMP-DOWN STATION REMOVED OUT-GASES FROM INSIDE THE TRIGGERS. (9/26/61) - Rocky Flats Plant, Final Assembly & Shipping, Eastern portion of plant site, south of Spruce Avenue, east of Tenth Street & north of Central Avenue, Golden, Jefferson County, CO

  16. WORKING PARK-FUEL CELL COMBINED HEAT AND POWER SYSTEM

    SciTech Connect

    Allan Jones

    2003-09-01

    This report covers the aims and objectives of the project which was to design, install and operate a fuel cell combined heat and power (CHP) system in Woking Park, the first fuel cell CHP system in the United Kingdom. The report also covers the benefits that were expected to accrue from the work in an understanding of the full technology procurement process (including planning, design, installation, operation and maintenance), the economic and environmental performance in comparison with both conventional UK fuel supply and conventional CHP and the commercial viability of fuel cell CHP energy supply in the new deregulated energy markets.

  17. Small Externally-fueled Heat Pipe Thermionic Reactor (SEHPTR) for dual mode applications

    NASA Astrophysics Data System (ADS)

    Malloy, John; Jacox, Michael; Zubrin, Robert

    1992-07-01

    The Small Externally Fueled Heat-Pipe Thermionic Reactor (SEHPTR) is described in the context of applications as a dual-mode nuclear power source for satellites. The SEHPTR is a thermionic power system based on a reactor with modular fuel elements around cylindrical thermionic heat-pipe modules with diodes for heat rejection. The SEHPTR concept is theorized to be suitable for directly heating hydrogen gas in the core to increase propulsion and reduce orbit-transfer times. The advantages of dual-mode operation of the SEHPTR are listed including enhanced mission safety and performance at relatively low costs. The SEHPTR could provide direct thermal thrust and an integrated stage that symbiotically utilizes electric power, direct thrust, and hydrogen arcjets. The system is argued to be more effective than a nuclear power system designed solely for electrical power production.

  18. Monticello BWR spent fuel assembly decay heat predictions and measurements

    SciTech Connect

    McKinnon, M.A.; Doman, J.W.; Heeb, C.M.; Creer, J.M.

    1986-06-01

    This report compares pre-calorimetry predictions of rates of six 7 x 7 boiling water reactor (BWR) spent fuel assemblies with measured decay heat rates. The assemblies were from Northern States Power Company's Monticello Nuclear Generating Plant and had burnups of 9 to 21 GWd/MTU and cooling times of 9 to 10 years. Conclusions are: The agreement between ORIGEN2 predictions and decay heat measurements of Monticello spent fuel is dependent on the method used to calibrate the calorimeter and to make the decay heat measurements. The agreement between predictions and measurements of decay heat rates of Monticello fuel is the same as that for Cooper and Dresden fuel if the same measurement method is used. The predictions are within a standard deviation of +-15 W of the measurements. Using a different measurement method, ORIGEN2 underpredicts the measured decay heat output of Monticello fuel assemblies by a constant 20 +- 2 W. The 20-W offset appears to be an artifact of the calibration procedure. The constant term in the calibration curve (i.e., q/sub DH/ = mx + b) can account for measurement differences of 40 W based on the 1983, 1984, and 1985 calibration curves. The difference between ORIGEN2 predictions and calorimeter decay heat measurements does not appear to be dependent on the magnitude of decay heat output. Predicted axial decay heat profiles are in good agreement with measured axial gamma radiation profiles. Recommendations are: Predictions using other decay heat codes should be compared to experimental data contained in this report, to evaluate prediction capabilities. The source of the differences that exist among calorimeter calibration curves needs to be determined. Calorimeter operational methods need to be investigated further to determine cause and effect relationships between operational method and calorimeter precision and accuracy.

  19. 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.

  20. 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)

  1. 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.

  2. Development of Passive Fuel Cell Thermal Management Heat Exchanger

    NASA Technical Reports Server (NTRS)

    Burke, Kenneth A.; Jakupca, Ian J.; Colozza, Anthony J.

    2010-01-01

    The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA Exploration program. The passive thermal management system relies on heat conduction within highly thermally conductive cooling plates to move the heat from the central portion of the cell stack out to the edges of the fuel cell stack. Using the passive approach eliminates the need for a coolant pump and other cooling loop components within the fuel cell system which reduces mass and improves overall system reliability. Previous development demonstrated the performance of suitable highly thermally conductive cooling plates that could conduct the heat, provide a sufficiently uniform temperature heat sink for each cell of the fuel cell stack, and be substantially lighter than the conventional thermal management approach. Tests were run with different materials to evaluate the design approach to a heat exchanger that could interface with the edges of the passive cooling plates. Measurements were made during fuel cell operation to determine the temperature of individual cooling plates and also to determine the temperature uniformity from one cooling plate to another.

  3. Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement

    SciTech Connect

    James A. Smith; Dale K. Kotter; Randall A. Alli; Steven L. Garrett

    2013-05-01

    A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

  4. Heat transfer in a fuel pin shipping container. [IDENT 1578

    SciTech Connect

    Ingham, J.G.

    1980-11-11

    Maximum cladding temperatures occur when the IDENT 1578 fuel pin shipping container is installed in the T-3 Cask. The maximum allowable cladding temperature of 800/sup 0/F is reached when the rate of energy deposited in the 19-pin basket reaches 400 watts. Since 45% of the energy which is generated in the fuel escapes the 19-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 400/.55 = 727 watts. Similarly, the maximum allowable cladding temperature of 800/sup 0/F is reached when the rate of energy deposited in the 40-pin basket reaches 465 watts. Since 33% of the energy which is generated in the fuel escapes the 40-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 465/.66 = 704 watts. The IDENT 1578 fuel pin shipping container therefore meets its thermal design criteria. IDENT 1578 can handle fuel pins with a decay heat load of 600 watts while maintaining the maximum fuel pin cladding temperature below 800/sup 0/F. The emissivities which were determined from the test results for the basket tubes and container are relatively low and correspond to new, shiny conditions. As the IDENT 1578 container is exposed to high temperatures for extended periods of time during the transportation of fuel pins, the emissivities will probably increase. This will result in reduced temperatures.

  5. Fuel quality issues in the oil heat industry

    SciTech Connect

    Litzke, Wai-Lin

    1992-12-01

    The quality of fuel oil plays an essential role in combustion performance and efficient operation of residential heating equipment. With the present concerns by the oil-heat industry of declining fuel-oil quality, a study was initiated to identify the factors that have brought about changes in the quality of distillate fuel. A background of information will be provided to the industry, which is necessary to deal with the problems relating to the fuel. The high needs for servicing heating equipment are usually the result of the poor handling characteristics of the fuel during cold weather, the buildup of dirt and water in storage tanks, and microbial growth. A discussion of how to deal with these problems is presented in this paper. The effectiveness of fuel additives to control these problems of quality is also covered to help users better understand the functions and limitations of chemical treatment. Test data have been collected which measure and compare changes in the properties of fuel using selected additives.

  6. Deposit formation and heat transfer in hydrocarbon rocket fuels

    NASA Technical Reports Server (NTRS)

    Giovanetti, A. J.; Spadaccini, L. J.; Szetela, E. J.

    1984-01-01

    An experimental research program was undertaken to investigate the thermal stability and heat transfer characteristics of several hydrocarbon fuels under conditions that simulate high-pressure, rocket engine cooling systems. The rates of carbon deposition in heated copper and nickel-plated copper tubes were determined for RP-1, propane, and natural gas using a continuous flow test apparatus which permitted independent variation and evaluation of the effect on deposit formation of wall temperature, fuel pressure, and fuel velocity. In addition, the effects of fuel additives and contaminants, cryogenic fuel temperatures, and extended duration testing with intermittent operation were examined. Corrosion of the copper tube surface was detected for all fuels tested; however, plating the insides of the tubes with nickel reduced deposit formation and eliminated corrosion in most cases. The lowest rates of carbon deposition were obtained for natural gas, and the highest rates were obtained for propane. Forced-convection heat transfer film coefficients were satisfactorily correlated using a Nusselt-Reynolds-Prandtl number equation for all the fuels tested.

  7. On mobile element transport in heated Abee. [chondrite thermal metamorphism

    NASA Technical Reports Server (NTRS)

    Ikramuddin, M.; Lipschutz, M. E.; Gibson, E. K., Jr.

    1979-01-01

    Abee chondrite samples were heated at 700 C for one week at 0.00001 to 0.001 atm Ne or at 0.00001 atm H2. Samples heated in Ne showed greater loss of Bi and Se and greater retention of Zn than those heated in H2. An inverse relationship between Zn retention and ambient Ne pressure was found. Seven trace elements (Ag, Co, Cs, Ga, In, Te, and Tl) were retained or lost to the same extent regardless of the heating conditions. Variations in the apparent activation energy for C above and below 700 C suggest that diffusive loss from different hosts and/or different mobile transport processes over the temperature range may have been in effect.

  8. Heat production in an Archean crustal profile and implications for heat flow and mobilization of heat-producing elements

    NASA Technical Reports Server (NTRS)

    Ashwal, L. D.; Morgan, P.; Kelley, S. A.; Percival, J. A.

    1987-01-01

    Concentrations of heat producing elements (Th, U, and K) in 58 samples representative of the main lithologies in a 100-km transect of the Superior Province of the Canadian Shield have been obtained. The relatively large variation in heat production found among the silicic plutonic rocks is shown to correlate with modal abundances of accessory minerals, and these variations are interpreted as premetamorphic. The present data suggest fundamental differences in crustal radioactivity distributions between granitic and more mafic terrains, and indicate that a previously determined apparently linear heat flow-heat production relationship for the Kapuskasing area does not relate to the distribution of heat production with depth.

  9. Analysis of a heat recirculating cooler for fuel gas sulfur removal in solid oxide fuel cells

    NASA Astrophysics Data System (ADS)

    Richards, Geo. A.; Berry, David A.; Freed, Adam

    When using conventional fossil fuels, most fuel cell systems require sulfur removal as part of their fuel processing. A novel approach to enable conventional sulfur removal in high-temperature fuel processing is presented. Using established principles from heat-recirculating combustors, it is suggested that high-temperature syngas can be momentarily cooled to conditions that would permit conventional sulfur removal to be carried out at relatively low temperatures. The recirculated heat is then used to heat the gas back to conditions that are minimally less than the original temperature. A model for evaluating the performance of this concept is presented, and calculations suggest that relative to fuel cell applications, reasonable physical dimensions can be expected in actual applications. For high-pressure syngas (i.e., coal gasification), the physical dimensions will rise with the operating pressure.

  10. Heat transfer correlations for kerosene fuels and mixtures and physical properties for Jet A fuel

    NASA Technical Reports Server (NTRS)

    Ackerman, G. H.; Faith, L. E.

    1972-01-01

    Heat transfer correlations are reported for conventional Jet A fuel for both laminar and turbulent flow in circular tubes. Correlations were developed for cooling in turbine engines, but have broader applications in petroleum and chemical processing, and other industrial applications.

  11. Sludge, fuel degradation and reducing fouling on heat exchangers

    SciTech Connect

    Butcher, T.; Litzke, Wai Lin; Krajewski, R.; Celebi, Y.

    1992-02-01

    Brookhaven National Laboratory, under contract to the US Department of Energy, operates an oil heat research primarily to lower energy consumption in the 12 million oil heated homes in the US. The program objectives include: Improve steady state efficiency of oil heating equipment, Improve seasonal efficiencies, Eliminate or minimize factors which tend to degrade system performance. This paper provides an overview of the status of three specific projects which fall under the above objectives. This includes our fuel quality project, oil appliance venting and a project addressing efficiency degradation due to soot fouling of heat exchangers.

  12. Finite element analysis of heat transport in a hydrothermal zone

    SciTech Connect

    Bixler, N.E.; Carrigan, C.R.

    1987-01-01

    Two-phase heat transport in the vicinity of a heated, subsurface zone is important for evaluation of nuclear waste repository design and estimation of geothermal energy recovery, as well as prediction of magma solidification rates. Finite element analyses of steady, two-phase, heat and mass transport have been performed to determine the relative importance of conduction and convection in a permeable medium adjacent to a hot, impermeable, vertical surface. The model includes the effects of liquid flow due to capillarity and buoyancy and vapor flow due to pressure gradients. Change of phase, with its associated latent heat effects, is also modeled. The mechanism of capillarity allows for the presence of two-phase zones, where both liquid and vapor can coexist, which has not been considered in previous investigations. The numerical method employs the standard Galerkin/finite element method, using eight-node, subparametric or isoparametric quadrilateral elements. In order to handle the extreme nonlinearities inherent in two-phase, nonisothermal, porous-flow problems, steady-state results are computed by integrating transients out to a long time (a method that is highly robust).

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

    DOEpatents

    Shore, Lawrence; Matlin, Ramail

    2011-03-08

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

  14. 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)

  15. 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

  16. Finite element computer model of microwave heated ceramics

    SciTech Connect

    Liqiu Zhou; Gang Liu; Jian Zhou

    1995-12-31

    In this paper, a 3-D finite element model to simulate the heating pattern during microwave sintering of ceramics in a TE{sub 10}{sup n} single mode rectangular cavity is described. A series of transient temperature profiles and heating rates of the ceramic cylinder and cubic sample were calculated versus different parameters such as thermal conductivity, dielectric loss factor, microwave power level, and microwave energy distribution. These numerical solutions may provide a better understanding of thermal runaway and solutions to microwave sintering of ceramics.

  17. 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.

  18. Fuel Cell Combined Heat and Power Commercial Demonstration

    SciTech Connect

    Brooks, Kriston P.; Makhmalbaf, Atefe

    2014-09-02

    This is the annual report for the Market Transformation project as required by DOE EERE's Fuel Cell Technologies Office. We have been provided with a specific format. It describes the work that was done in developing evaluating the performance of 5 kW stationary combined heat and power fuel cell systems that have been deployed in Oregon and California. It also describes the business case that was developed to identify markets and address cost.

  19. Benchmark Wall Heat Flux Data for a GO2/GH2 Single Element Combustor

    NASA Technical Reports Server (NTRS)

    Marshall, William M.; Pal, Sibtosh; Woodward, Roger d.; Santoro, Robert J.

    2005-01-01

    Wall heat flux measurements in a 1.5 in. diameter circular cross-section rocket chamber for a uni-element shear coaxial injector element operating on gaseous oxygen (GOz)/gaseous hydrogen (GH,) propellants are presented. The wall heat flux measurements were made using arrays of Gardon type heat flux gauges and coaxial thermocouple instrumentation. Wall heat flux measurements were made for two cases. For the first case, GOZ/GHz oxidizer-rich (O/F=l65) and fuel-rich preburners (O/F=1.09) integrated with the main chamber were utilized to provide vitiated hot fuel and oxidizer to the study shear coaxial injector element. For the second case, the preburners were removed and ambient temperature gaseous oxygen/gaseous hydrogen propellants were supplied to the study injector. Experiments were conducted at four chamber pressures of 750, 600, 450 and 300psia for each case. The overall mixture ratio for the preburner case was 6.6, whereas for the ambient propellant case, the mixture ratio was 6.0. Total propellant flow was nominally 0.27-0.29 Ibm/s for the 750 psia case with flowrates scaled down linearly for lower chamber pressures. The axial heat flux profile results for both the preburner and ambient propellant cases show peak heat flux levels a t axial locations between 2.0 and 3.0 in. from the injector face. The maximum heat flux level was about two times greater for the preburner case. This is attributed to the higher injector fuel-to-oxidizer momentum flux ratio that promotes mixing and higher initial propellant temperature for the preburner case which results in a shorter reaction zone. The axial heat flux profiles were also scaled with respect to the chamber pressure to the power 0.8. The results at the four chamber pressures for both cases collapsed to a single profile indicating that at least to first approximation, the basic fluid dynamic structures in the flow field are pressure independent as long as the chamber/njector/nozzle geometry and injection velocities

  20. Iron aluminide useful as electrical resistance heating elements

    SciTech Connect

    Sikka, V.K.; Deevi, S.C.; Fleischhauer, G.S.; Hajaligol, M.R.; Lilly, A.C. Jr.

    1999-11-02

    The invention relates generally to aluminum containing iron-base alloys useful as electrical resistance heating elements. The aluminum containing iron-base alloys have improved room temperature ductility, electrical resistivity, cyclic fatigue resistance, high temperature oxidation resistance, low and high temperature strength, and/or resistance to high temperature sagging. The alloy has an entirely ferritic microstructure which is free of austenite and includes, in weight %, over 4% Al, {le}1% Cr and either {ge}0.05% Zr or ZrO{sub 2} stringers extending perpendicular to an exposed surface of the heating element or {ge}0.1% oxide dispersoid particles. The alloy can contain 14--32% Al, {le}2% Ti, {le}2% Mo, {le}1% Zr, {le}1% C, {le}0.1% B, {le}30% oxide dispersoid and/or electrically insulating or electrically conductive covalent ceramic particles, {le}1% rare earth metal, {le}1% oxygen, {le}3% Cu, balance Fe.

  1. Iron aluminide useful as electrical resistance heating elements

    DOEpatents

    Sikka, Vinod K.; Deevi, Seetharama C.; Fleischhauer, Grier S.; Hajaligol, Mohammad R.; Lilly, Jr., A. Clifton

    2001-01-01

    The invention relates generally to aluminum containing iron-base alloys useful as electrical resistance heating elements. The aluminum containing iron-base alloys have improved room temperature ductility, electrical resistivity, cyclic fatigue resistance, high temperature oxidation resistance, low and high temperature strength, and/or resistance to high temperature sagging. The alloy has an entirely ferritic microstructure which is free of austenite and includes, in weight %, over 4% Al, .ltoreq.1% Cr and either .gtoreq.0.05% Zr or ZrO.sub.2 stringers extending perpendicular to an exposed surface of the heating element or .gtoreq.0.1% oxide dispersoid particles. The alloy can contain 14-32% Al, .ltoreq.2% Ti, .ltoreq.2% Mo, .ltoreq.1% Zr, .ltoreq.1% C, .ltoreq.0.1% B, .ltoreq.30% oxide dispersoid and/or electrically insulating or electrically conductive covalent ceramic particles, .ltoreq.1% rare earth metal, .ltoreq.1% oxygen, .ltoreq.3% Cu, balance Fe.

  2. Iron aluminide useful as electrical resistance heating elements

    DOEpatents

    Sikka, Vinod K.; Deevi, Seetharama C.; Fleischhauer, Grier S.; Hajaligol, Mohammad R.; Lilly, Jr., A. Clifton

    1997-01-01

    The invention relates generally to aluminum containing iron-base alloys useful as electrical resistance heating elements. The aluminum containing iron-base alloys have improved room temperature ductility, electrical resistivity, cyclic fatigue resistance, high temperature oxidation resistance, low and high temperature strength, and/or resistance to high temperature sagging. The alloy has an entirely ferritic microstructure which is free of austenite and includes, in weight %, over 4% Al, .ltoreq.1% Cr and either .gtoreq.0.05% Zr or ZrO.sub.2 stringers extending perpendicular to an exposed surface of the heating element or .gtoreq.0.1% oxide dispersoid particles. The alloy can contain 14-32% Al, .ltoreq.2% Ti, .ltoreq.2% Mo, .ltoreq.1% Zr, .ltoreq.1% C, .ltoreq.0.1% B, .ltoreq.30% oxide dispersoid and/or electrically insulating or electrically conductive covalent ceramic particles, .ltoreq.1% rare earth metal, .ltoreq.1% oxygen, .ltoreq.3% Cu, balance Fe.

  3. Iron aluminide useful as electrical resistance heating elements

    DOEpatents

    Sikka, V.K.; Deevi, S.C.; Fleischhauer, G.S.; Hajaligol, M.R.; Lilly, A.C. Jr.

    1997-04-15

    The invention relates generally to aluminum containing iron-base alloys useful as electrical resistance heating elements. The aluminum containing iron-base alloys have improved room temperature ductility, electrical resistivity, cyclic fatigue resistance, high temperature oxidation resistance, low and high temperature strength, and/or resistance to high temperature sagging. The alloy has an entirely ferritic microstructure which is free of austenite and includes, in weight %, over 4% Al, {<=}1% Cr and either {>=}0.05% Zr or ZrO{sub 2} stringers extending perpendicular to an exposed surface of the heating element or {>=}0.1% oxide dispersoid particles. The alloy can contain 14-32% Al, {<=}2% Ti, {<=}2% Mo, {<=}1% Zr, {<=}1% C, {<=}0.1% B, {<=}30% oxide dispersoid and/or electrically insulating or electrically conductive covalent ceramic particles, {<=}1% rare earth metal, {<=}1% oxygen, {<=}3% Cu, balance Fe. 64 figs.

  4. Iron aluminide useful as electrical resistance heating elements

    DOEpatents

    Sikka, Vinod K.; Deevi, Seetharama C.; Fleischhauer, Grier S.; Hajaligol, Mohammad R.; Lilly, Jr., A. Clifton

    1999-01-01

    The invention relates generally to aluminum containing iron-base alloys useful as electrical resistance heating elements. The aluminum containing iron-base alloys have improved room temperature ductility, electrical resistivity, cyclic fatigue resistance, high temperature oxidation resistance, low and high temperature strength, and/or resistance to high temperature sagging. The alloy has an entirely ferritic microstructure which is free of austenite and includes, in weight %, over 4% Al, .ltoreq.1% Cr and either .gtoreq.0.05% Zr or ZrO.sub.2 stringers extending perpendicular to an exposed surface of the heating element or .gtoreq.0.1% oxide dispersoid particles. The alloy can contain 14-32% Al, .ltoreq.2% Ti, .ltoreq.2% Mo, .ltoreq.1% Zr, .ltoreq.1% C, .ltoreq.0.1% B, .ltoreq.30% oxide dispersoid and/or electrically insulating or electrically conductive covalent ceramic particles, .ltoreq.1% rare earth metal, .ltoreq.1% oxygen, .ltoreq.3% Cu, balance Fe.

  5. Heat recovery and pollutant cleanup from low grade fuels

    SciTech Connect

    Ellison, W.; Butcher, T.A.; Carbonara, J.C.; Heaphy, J.P.

    1994-06-01

    Technical development efforts and field testing have pointed to outstanding economy and environmental benefits contemplated in revamping of fueling for reduced cost of power generation. Flue gas cleaning technologies detailed herein are expected to vitally support this objective and strongly contribute to long-term efforts for regional ozone compliance within the favorable economic framework made possible by avoidance of clean, high-cost, steam boiler fuels otherwise necessary in meeting environmental goals. With adequate control of emissions, abundance and attractive price of high-sulfur residium or coal provides the realistic basis for cost-effective power generation in decades ahead. A key element is the design of by-product yielding, wet flue gas desulfurization processes. The choice is among those using lime, ammonia, or sodium alkali reagents, or limestone in highly oxygen-inhibited process operation, with SO{sub 2} removal efficiency of 98+% as a result of dissolved sulfite alkalinity. Integrated use of condensing heat exchangers provides low-level heat recovery and water-condensing-mode scrubbing. SO{sub 3} gas & PM-10 particulates including trace metals are effectively removed in conjunction with optimal, ultra-efficient, simultaneous multi-pollutant reduction. DeNO{sub x} may be accomplished by combining advantageous recirculation of highly-cooled, low-humidity, clean flue gas to burner windboxes with conventional selective non-catalytic reduction. Stack NO{sub x} at 18 to 30 ppM, (60% O{sub 2} basis), i.e. 0.03 to 0.05 lb NO{sub 2}-equivalent/MM Btu, may be achieved by injection of methanol in dilute solution or highly air-diluted, into the rear boiler cavity upstream of the economizer, converting flue-gas NO to NO{sub 2}, thereafter efficiently absorbed and chemically reduced to N{sub 2} by the dissolved-sulfite scrubbing agent to gain colorless discharge with NO{sub 2} concentration less than 15 ppM, i.e. 0.025 lb/MM Btu.

  6. Method for producing heat treated composite nuclear fuel containers

    SciTech Connect

    Rosenbaum, H.S.

    1993-06-29

    A method is described for producing composite constructed nuclear fuel containers for service in water cooled nuclear fission reactors comprising a tubular zirconium alloy casing having a protective lining of a zirconium metal covering the inside surface of the zirconium alloy tubular casing and being metallurgically bonded thereto, consisting essentially of the steps of: heat treating a large diameter zirconium alloy tube comprising a beta-quench treatment of heating the zirconium alloy to a temperature sufficient to recrystallize the zirconium alloy to its beta phase of at least about 970 C and then rapidly cooling the thus heated and recrystallized zirconium alloy tube stock, heat treating a large diameter zirconium metal hollow liner stock comprising a beta-quench treatment of heating the zirconium metal to a temperature sufficient to recrystallize the zirconium metal to its beta phase of at least about 900 C and then rapidly cooling the thus heated and recrystallized zirconium metal hollow liner stock, assembling the heat treated zirconium alloy tube stock and heat treated lining stock with the hollow zirconium metal lining stock inserted inclose fitting contact within the tube stock, metallurgically bonding the tube and liner stocks providing a composite tubular stock unit with the tube stock surrounding the lining stock, and reducing the circumference of the assembled composite tubular stock unit down to a size suitable for service as a lined tubular container for nuclear fuel in a series of progressive reductions in circumference applied in sequence.

  7. TECHNICAL NOTE: A feasibility study of self-heating concrete utilizing carbon nanofiber heating elements

    NASA Astrophysics Data System (ADS)

    Chang, Christiana; Ho, Michelle; Song, Gangbing; Mo, Yi-Lung; Li, Hui

    2009-12-01

    This paper presents the development of an electric, self-heating concrete system that uses embedded carbon nanofiber paper as electric resistance heating elements. The proposed system utilizes the conductive properties of carbon fiber materials to heat a surface overlay of concrete with various admixtures to improve the concrete's thermal conductivity. The development and laboratory scale testing of the system were conducted for the various compositions of concrete containing, separately, carbon fiber, fly ash, and steel shavings as admixtures. The heating performances of these concrete mixtures with the carbon fiber heating element were experimentally obtained in a sub-freezing ambient environment in order to explore the use of such a system for deicing of concrete roadways. Analysis of electric power consumption, heating rate, and obtainable concrete surface temperatures under typical power loads was performed to evaluate the viability of a large scale implementation of the proposed heating system for roadway deicing applications. A cost analysis is presented to provide a comparison with traditional deicing methods, such as salting, and other integrated concrete heating systems.

  8. Solar Thermochemical Fuels Production: Solar Fuels via Partial Redox Cycles with Heat Recovery

    SciTech Connect

    2011-12-19

    HEATS Project: The University of Minnesota is developing a solar thermochemical reactor that will efficiently produce fuel from sunlight, using solar energy to produce heat to break chemical bonds. The University of Minnesota is envisioning producing the fuel by using partial redox cycles and ceria-based reactive materials. The team will achieve unprecedented solar-to-fuel conversion efficiencies of more than 10% (where current state-of-the-art efficiency is 1%) by combined efforts and innovations in material development, and reactor design with effective heat recovery mechanisms and demonstration. This new technology will allow for the effective use of vast domestic solar resources to produce precursors to synthetic fuels that could replace gasoline.

  9. Application of the boundary element method to transient heat conduction

    NASA Technical Reports Server (NTRS)

    Dargush, G. F.; Banerjee, P. K.

    1991-01-01

    An advanced boundary element method (BEM) is presented for the transient heat conduction analysis of engineering components. The numerical implementation necessarily includes higher-order conforming elements, self-adaptive integration and a multiregion capability. Planar, three-dimensional and axisymmetric analyses are all addressed with a consistent time-domain convolution approach, which completely eliminates the need for volume discretization for most practical analyses. The resulting general purpose algorithm establishes BEM as an attractive alternative to the more familiar finite difference and finite element methods for this class of problems. Several detailed numerical examples are included to emphasize the accuracy, stability and generality of the present BEM. Furthermore, a new efficient treatment is introduced for bodies with embedded holes. This development provides a powerful analytical tool for transient solutions of components, such as casting moulds and turbine blades, which are cumbersome to model when employing the conventional domain-based methods.

  10. On the heat capacity of elements in WMD regime

    NASA Astrophysics Data System (ADS)

    Hamel, Sebatien

    2014-03-01

    Once thought to get simpler with increasing pressure, elemental systems have been discovered to exhibit complex structures and multiple phases at high pressure. For carbon, QMD/PIMC simulations have been performed and the results are guiding alternative modelling methodologies for constructing a carbon equation-of-state covering the warm dense matter regime. One of the main results of our new QMD/PIMC carbon equation of state is that the decay of the ion-thermal specific heat with temperature is much faster than previously expected. An important question is whether this is only found in carbon and not other element. In this presentation, based on QMD calculations for several elements, we explore trends in the transition from condensed matter to warm dense matter regime.

  11. Conjugate Heat Transfer Analyses on the Manifold for Ramjet Fuel Injectors

    NASA Technical Reports Server (NTRS)

    Wang, Xiao-Yen J.

    2006-01-01

    Three-dimensional conjugate heat transfer analyses on the manifold located upstream of the ramjet fuel injector are performed using CFdesign, a finite-element computational fluid dynamics (CFD) software. The flow field of the hot fuel (JP-7) flowing through the manifold is simulated and the wall temperature of the manifold is computed. The three-dimensional numerical results of the fuel temperature are compared with those obtained using a one-dimensional analysis based on empirical equations, and they showed a good agreement. The numerical results revealed that it takes around 30 to 40 sec to reach the equilibrium where the fuel temperature has dropped about 3 F from the inlet to the exit of the manifold.

  12. Deterioration in heat transfer of endothermal hydrocarbon fuel

    NASA Astrophysics Data System (ADS)

    Zhou, Weixing; Bao, Wen; Qin, Jiang; Qu, Yunfeng

    2011-06-01

    Numerical studies under supercritical pressure are carried out to study the heat transfer characteristics in a single-root coolant channel of the active regenerative cooling system of the scramjet engine, using actual physical properties of pentane. The relationships between wall temperature and inlet temperature, mass flow rate, wall heat flux, inlet pressure, as well as center stream temperature are obtained. The results suggest that the heat transfer deterioration occurs when the fuel temperature approaches the pseudo-critical temperature, and the wall temperature increases rapidly and heat transfer coefficient decreases sharply. The decrease of wall heat flux, as well as the increase of mass flow rate and inlet pressure makes the starting point of the heat transfer deterioration and the peak point of the wall temperature move backward. The wall temperature increment induced by heat transfer deterioration decreases, which could reduce the severity of the heat transfer deterioration. The relational expression of the heat transfer deterioration critical heat flux derives from the relationship of the mass flow rate and the inlet pressure.

  13. Heat Transfer Modeling of Dry Spent Nuclear Fuel Storage Facilities

    SciTech Connect

    Lee, S.Y.

    1999-01-13

    The present work was undertaken to provide heat transfer model that accurately predicts the thermal performance of dry spent nuclear fuel storage facilities. One of the storage configurations being considered for DOE Aluminum-clad Spent Nuclear Fuel (Al-SNF), such as the Material and Testing Reactor (MTR) fuel, is in a dry storage facility. To support design studies of storage options a computational and experimental program has been conducted at the Savannah River Site (SRS). The main objective is to develop heat transfer models including natural convection effects internal to an interim dry storage canister and to geological codisposal Waste Package (WP). Calculated temperatures will be used to demonstrate engineering viability of a dry storage option in enclosed interim storage and geological repository WP and to assess the chemical and physical behaviors of the Al-SNF in the dry storage facilities. The current paper describes the modeling approaches and presents the computational results along with the experimental data.

  14. Potential of vegetable oils as a domestic heating fuel

    SciTech Connect

    Hayden, A.C.S.; Begin, E.; Palmer, C.E.

    1982-06-01

    The dependence on imported oil for domestic heating has led to the examination of other potential fuel substitutes. One potential fuel is some form of vegetable oil, which could be a yearly-renewable fuel. In Western Canada, canola has become a major oilseed crop; in Eastern Canada, sunflowers increasingly are becoming a source for a similar oil; for this reason, the Canadian Combustion Research Laboratory (CCRL) has chosen these oils for experimentation. Trials have been conducted in a conventional warm air oil furnace, fitted with a flame retention head burner. Performance has been measured with pure vegetable oils as well as a series of blends with conventional No. 2 oil. The effects of increased fuel pressure and fuel preheating are established. Emissions of carbon monoxide, nitrogen oxides, unburned hydrocarbons and particulates are given for both steady state and cyclic operation. Canola oil cannot be fired in cyclic operation above 50:50 blends with No. 2 oil. At any level above a 10% blend, canola is difficult to burn, even with significant increased pressure and temperature. Sunflower oil is much easier to burn and can be fired as a pure fuel, but with high emissions of incomplete combustion products. An optimum blend of 50:50 sunflower in No. 2 oil yields emissions and performance similar to No. 2 oil. This blend offers potential as a means of reducing demand of imported crude oil for domestic heating systems.

  15. Radionuclide mass inventory, activity, decay heat, and dose rate parametric data for TRIGA spent nuclear fuels

    SciTech Connect

    Sterbentz, J.W.

    1997-03-01

    Parametric burnup calculations are performed to estimate radionuclide isotopic mass and activity concentrations for four different Training, Research, and Isotope General Atomics (TRIGA) nuclear reactor fuel element types: (1) Aluminum-clad standard, (2) Stainless Steel-clad standard, (3) High-enrichment Fuel Life Improvement Program (FLIP), and (4) Low-enrichment Fuel Life Improvement Program (FLIP-LEU-1). Parametric activity data are tabulated for 145 important radionuclides that can be used to generate gamma-ray emission source terms or provide mass quantity estimates as a function of decay time. Fuel element decay heats and dose rates are also presented parametrically as a function of burnup and decay time. Dose rates are given at the fuel element midplane for contact, 3.0-feet, and 3.0-meter detector locations in air. The data herein are estimates based on specially derived Beginning-of-Life (BOL) neutron cross sections using geometrically-explicit TRIGA reactor core models. The calculated parametric data should represent good estimates relative to actual values, although no experimental data were available for direct comparison and validation. However, because the cross sections were not updated as a function of burnup, the actinide concentrations may deviate from the actual values at the higher burnups.

  16. AERIAL MEASUREMENTS OF CONVECTION CELL ELEMENTS IN HEATED LAKES

    SciTech Connect

    Villa-Aleman, E; Saleem Salaymeh, S; Timothy Brown, T; Alfred Garrett, A; Malcolm Pendergast, M; Linda Nichols, L

    2007-12-19

    Power plant-heated lakes are characterized by a temperature gradient in the thermal plume originating at the discharge of the power plant and terminating at the water intake. The maximum water temperature discharged by the power plant into the lake depends on the power generated at the facility and environmental regulations on the temperature of the lake. Besides the observed thermal plume, cloud-like thermal cells (convection cell elements) are also observed on the water surface. The size, shape and temperature of the convection cell elements depends on several parameters such as the lake water temperature, wind speed, surfactants and the depth of the thermocline. The Savannah River National Laboratory (SRNL) and Clemson University are collaborating to determine the applicability of laboratory empirical correlations between surface heat flux and thermal convection intensity. Laboratory experiments at Clemson University have demonstrated a simple relationship between the surface heat flux and the standard deviation of temperature fluctuations. Similar results were observed in the aerial thermal imagery SRNL collected at different locations along the thermal plume and at different elevations. SRNL will present evidence that the results at Clemson University are applicable to cooling lakes.

  17. Aerial measurements of convection cell elements in heated lakes

    NASA Astrophysics Data System (ADS)

    Villa-Aleman, E.; Salaymeh, S. R.; Brown, T. B.; Garrett, A. J.; Nichols, L. S.; Pendergast, M. M.

    2008-03-01

    Power plant-heated lakes are characterized by a temperature gradient in the thermal plume originating at the discharge of the power plant and terminating at the water intake. The maximum water temperature discharged by the power plant into the lake depends on the power generated at the facility and environmental regulations on the temperature of the lake. Besides the observed thermal plume, cloud-like thermal cells (convection cell elements) are also observed on the water surface. The size, shape and temperature of the convection cell elements depends on several parameters such as the lake water temperature, wind speed, surfactants and the depth of the thermocline. The Savannah River National Laboratory (SRNL) and Clemson University are collaborating to determine the applicability of laboratory empirical correlations between surface heat flux and thermal convection intensity. Laboratory experiments at Clemson University have demonstrated a simple relationship between the surface heat flux and the standard deviation of temperature fluctuations. Similar results were observed in the aerial thermal imagery SRNL collected at different locations along the thermal plume and at different elevations. SRNL will present evidence that the results at Clemson University are applicable to cooling lakes.

  18. Specific features of external heat and mass transfer in the vibration apparatuses used for regenerating spent fuel from nuclear power plants

    NASA Astrophysics Data System (ADS)

    Sapozhnikov, B. G.; Gorbunova, A. M.; Zelenkova, Yu. O.; Sapozhnikov, G. B.; Shiryaeva, N. P.

    2014-06-01

    We present experimental data on the coefficients of heat and mass transfer for freely floating bodies simulating fragments of cladding and large conglomerates of fuel, as well as on the local coefficients of heat and mass transfer over the bed height, which point to high intensity of heat and mass transfer processes that take place in the elements of vibration apparatuses intended for subjecting spent fuel from nuclear power plants to oxidative recrystallization.

  19. 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.

  20. HIFU Induced Heating Modelling by Using the Finite Element Method

    NASA Astrophysics Data System (ADS)

    Martínez, R.; Vera, A.; Leija, L.

    High intensity focused ultrasound is a thermal therapy method used to treat malignant tumors and other medical conditions. Focused ultrasound concentrates acoustic energy at a focal zone. There, temperature rises rapidly over 56 °C to provoke tissue necrosis. Device performance depends on its fabrication placing computational modeling as a powerful tool to anticipate experimentation results. Finite element method allows modeling of multiphysics systems. Therefore, induced heating was modeled considering the acoustic field produced by a concave radiator excited with electric potentials from 5 V to 20 V. Nonlinear propagation was neglected and a linear response between the acoustic fields and pressure distribution was obtained. Finally, the results showed that acoustic propagation and heating models should be improved and validated with experimental measurements.

  1. Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

    NASA Astrophysics Data System (ADS)

    Najeeb, Umair

    This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

  2. 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.

  3. Temperature characteristics for PTC material heating diesel fuel

    NASA Astrophysics Data System (ADS)

    Gu, Lefeng; Li, Xiaolu; Wang, Jun; Li, Ying; Li, Ming

    2010-08-01

    This paper gives a way which utilizes the PTC (Positive Temperature Coefficient) material to preheat diesel fuel in the injector in order to improve the cold starting and emissions of engine. A new injector is also designed. In order to understand the preheating process in this new injector, a dynamic temperature testing system combined with the MSP430F149 data acquisition system is developed for PTC material heating diesel fuel. Especially, the corresponding software and hardware circuits are explained. The temperature of diesel fuel preheating by PTC ceramics is measured under different voltages and distances, which Curie point is 75 °C. Diesel fuel is heated by self-defined temperature around the Curie point of PTC ceramics. The diesel fuel temperature rises rapidly in 2 minutes of the beginning, then can reach 60 °C within 5 minutes as its distance is 5mm away from the surface of PTC ceramics. However, there are a lot of fundamental studies and technology to be resolved in order to apply PTC material in the injector successfully.

  4. 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

  5. Utility reduces fuel cost with heat recovery, industrial byproduct fuel, cogeneration

    SciTech Connect

    Holland, R.J.

    1982-02-01

    A 50-MW North Dakota power plant is refurbished to recover major waste-heat sources. Use of agricultural byproduct fuel and cogeneration also helps to cut future costs. The plant is saving on fuel costs by burning 150-200 tons/day of sunflower seed hulls from a local processing plant. The hulls are pulverized and mixed with the primary fuel, North Dakota lignite. At the same time, the processing plant that supplies the sunflower hulls buys steam from the power plant, thus giving the utility some of the economic benefits of cogeneration.

  6. TACO3D. 3-D Finite Element Heat Transfer Code

    SciTech Connect

    Mason, W.E.

    1992-03-04

    TACO3D is a three-dimensional, finite-element program for heat transfer analysis. An extension of the two-dimensional TACO program, it can perform linear and nonlinear analyses and can be used to solve either transient or steady-state problems. The program accepts time-dependent or temperature-dependent material properties, and materials may be isotropic or orthotropic. A variety of time-dependent and temperature-dependent boundary conditions and loadings are available including temperature, flux, convection, and radiation boundary conditions and internal heat generation. Additional specialized features treat enclosure radiation, bulk nodes, and master/slave internal surface conditions (e.g., contact resistance). Data input via a free-field format is provided. A user subprogram feature allows for any type of functional representation of any independent variable. A profile (bandwidth) minimization option is available. The code is limited to implicit time integration for transient solutions. TACO3D has no general mesh generation capability. Rows of evenly-spaced nodes and rows of sequential elements may be generated, but the program relies on separate mesh generators for complex zoning. TACO3D does not have the ability to calculate view factors internally. Graphical representation of data in the form of time history and spatial plots is provided through links to the POSTACO and GRAPE postprocessor codes.

  7. Utilization of waste heat in trucks for increased fuel economy

    NASA Technical Reports Server (NTRS)

    Leising, C. J.; Purohit, G. P.; Degrey, S. P.; Finegold, J. G.

    1978-01-01

    Improvements in fuel economy for a broad spectrum of truck engines and waste heat utilization concepts are evaluated and compared. The engines considered are the diesel, spark ignition, gas turbine, and Stirling. The waste heat utilization concepts include preheating, regeneration, turbocharging, turbocompounding, and Rankine engine compounding. Predictions were based on fuel-air cycle analyses, computer simulation, and engine test data. The results reveal that diesel driving cycle performance can be increased by 20% through increased turbocharging, turbocompounding, and Rankine engine compounding. The Rankine engine compounding provides about three times as much improvement as turbocompounding but also costs about three times as much. Performance for either is approximately doubled if applied to an adiabatic diesel.

  8. Heating experiments for flowability improvement of near-freezing aviation fuel

    NASA Technical Reports Server (NTRS)

    Friedman, R.; Stockemer, F. J.

    1984-01-01

    An experimental jet fuel with a -33 C freezing point was chilled in a wing tank simulator with superimposed fuel heating to improve low temperature flowability. Heating consisted of circulating a portion of the fuel to an external heat exchanger and returning the heated fuel to the tank. Flowability was determined by the mass percent of unpumpable fuel (holdup) left in the simulator upon withdrawal of fuel at the conclusion of testing. The study demonstrated that fuel heating is feasible and improves flowability as compared to that of baseline, unheated tests. Delayed heating with initiation when the fuel reaches a prescribed low temperature limit, showed promise of being more efficient than continuous heating. Regardless of the mode or rate of heating, complete flowability (zero holdup) could not be restored by fuel heating. The severe, extreme-day environment imposed by the test caused a very small amount of subfreezing fuel to be retained near the tank surfaces even at high rates of heating. Correlations of flowability established for unheated fuel tests also could be applied to the heated test results if based on boundary-layer temperature or a solid index (subfreezing point) characteristic of the fuel. Previously announced in STAR as N82-26483

  9. 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.

  10. Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses

    SciTech Connect

    Rector, D.R.; McCann, R.A.; Jenquin, U.P.; Heeb, C.M.; Creer, J.M.; Wheeler, C.L.

    1986-12-01

    This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions.

  11. Distribution of heating in an LVRF bundle due to dysprosium in the central element

    SciTech Connect

    Tsang, K.; Buijs, A.

    2006-07-01

    The computer code MCNP was used to establish the effect of adding dysprosium to the central pin of the proposed BRUCE-B CANFLEX{sup R} Low-Void-Reactivity Fuel (LVRF) on the heat load of the central pin and the heat balance inside the fuel bundle. The Dy generates heat through radiative capture of thermal neutrons, as well as through beta decay of {sup 165}Dy to {sup 165}Ho. We conclude that for fresh fuel, the presence of Dy contributes 26% of the overall heat to the central pin, and 0.5% to the whole fuel bundle. These percentages decrease to 11% and 0.5% at the end-of-life burnup condition. A second, operational quantity is the HPFP ratio (heating-power to fission-power ratio). This ratio is 1.63 for fresh fuel and decreases to 1.19 for fuel at the end-of-life burnup condition. (authors)

  12. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    SciTech Connect

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  13. Maine State Planning Office, 1990--1991 heating season home heating fuels price survey

    SciTech Connect

    Not Available

    1991-01-01

    The 1990--1991 heating season was the first time in Maine that the Home Heating Fuels Survey was conducted for the United States Department of Energy by the Maine State Planning Office. This season also marked the first time that dealers were surveyed for a price for propane. Under a late agreement, the State of Maine was picked up by the regional survey of the Energy Information Agency in the beginning of October. This accounted for the weekly survey of the traditional participants in the State's Home Heating Fuels Price Survey being supplemented by biweekly DOE surveys of separate survey samples of oil and propane dealers. The SPO sample identifies 36 dealers in the State of Maine, while the DOE sample was constructed around 22 oil dealers in Maine and New Hampshire and 29 propane dealers in Maine.

  14. Maine State Planning Office, 1990--1991 heating season home heating fuels price survey. Final report

    SciTech Connect

    Not Available

    1991-12-31

    The 1990--1991 heating season was the first time in Maine that the Home Heating Fuels Survey was conducted for the United States Department of Energy by the Maine State Planning Office. This season also marked the first time that dealers were surveyed for a price for propane. Under a late agreement, the State of Maine was picked up by the regional survey of the Energy Information Agency in the beginning of October. This accounted for the weekly survey of the traditional participants in the State`s Home Heating Fuels Price Survey being supplemented by biweekly DOE surveys of separate survey samples of oil and propane dealers. The SPO sample identifies 36 dealers in the State of Maine, while the DOE sample was constructed around 22 oil dealers in Maine and New Hampshire and 29 propane dealers in Maine.

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

    NASA Astrophysics Data System (ADS)

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

    2001-09-01

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

  16. 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.

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

    SciTech Connect

    Not Available

    1994-06-01

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

  18. 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.

  19. Elemental characteristics of aerosols emitted from a coal-fired heating plant

    NASA Technical Reports Server (NTRS)

    Singh, J. J.; Khandelwal, G. S.

    1978-01-01

    Size differentiated aerosols were collected downstream from a heating plant fueled with eastern coal and analyzed using particle induced X-ray emission technique. Based on aerosol masses collected in various size ranges, the aerosol size distribution is determined to be trimodal, with the three peaks centered at 0.54 microns, 4.0 microns, and 11.0 microns, respectively. Of the various trace elements present in the aerosols, sulphur is the only element that shows very strong concentration in the smallest size group. Iron is strongly concentrated in the 4.0 micron group. Potassium, calcium, and titanium also exhibit stronger concentration in the 4.0 micron group than any other group. Other trace elements - vanadium, chromium, manganese, nickel, copper, and barium - are equally divided between the 0.54 microns and the 4.0 microns groups. Apparently, all of the trace elements - except S - enter aerosols during the initial formation and subsequent condensation phases in the combustion process. Excess concentration of sulphur in the 0.54 microns group can only be accounted for by recondensation of sulphur vapors on the combustion aerosols and gas-to-particle phase conversion of sulfate vapors at the stack top.

  20. Performance gains by using heated natural-gas fuel in an annular turbojet combustor

    NASA Technical Reports Server (NTRS)

    Marchionna, N. R.

    1973-01-01

    A full-scale annular turbojet combustor was tested with natural gas fuel heated from ambient temperature to 800 K (980 F). In all tests, heating the fuel improved combustion efficiency. Two sets of gaseous fuel nozzles were tested. Combustion instabilities occurred with one set of nozzles at two conditions: one where the efficiency approached 100 percent with the heated fuel; the other where the efficiency was very poor with the unheated fuel. The second set of nozzles exhibited no combustion instability. Altitude relight tests with the second set showed that relight was improved and was achievable at essentially the same condition as blowout when the fuel temperature was 800 K (980 F).

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

    SciTech Connect

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

    2007-09-01

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

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

    SciTech Connect

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

    1999-02-01

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

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

    SciTech Connect

    Maassen, D. P.

    1997-10-15

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

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

    SciTech Connect

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

    1992-08-01

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

  5. 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.

  6. Fission product retention in TRISCO coated UO sub 2 particle fuels subjected to HTR simulated core heating tests

    SciTech Connect

    Baldwin, C.A.; Kania, M.J.

    1990-11-01

    Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600{degree}C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800{degree}C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800{degree}C and above may exist. 6 refs., 6 figs., 4 tabs.

  7. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    SciTech Connect

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  8. 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.

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

    SciTech Connect

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

    1990-07-01

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

  10. Heat transfer in hydrocarbon fuel boiling under conditions of natural convection

    SciTech Connect

    Shigabiev, T.N.; Galimov, F.M.

    1995-12-01

    Data on the heat-transfer coefficient in boiling of five jet fuels, two automotive gasolines, and a diesel fuel are presented over a wide range of regime parameters. The obtained results are described by a unified similarity equation.

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

    SciTech Connect

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

    1999-07-26

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

  12. ZPPR FUEL ELEMENT THERMAL STRESS-STRAIN ANALYSIS

    SciTech Connect

    Charles W. Solbrig; Jason Andrus; Chad Pope

    2014-04-01

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

  13. Radiation Heat Transfer Between Diffuse-Gray Surfaces Using Higher Order Finite Elements

    NASA Technical Reports Server (NTRS)

    Gould, Dana C.

    2000-01-01

    This paper presents recent work on developing methods for analyzing radiation heat transfer between diffuse-gray surfaces using p-version finite elements. The work was motivated by a thermal analysis of a High Speed Civil Transport (HSCT) wing structure which showed the importance of radiation heat transfer throughout the structure. The analysis also showed that refining the finite element mesh to accurately capture the temperature distribution on the internal structure led to very large meshes with unacceptably long execution times. Traditional methods for calculating surface-to-surface radiation are based on assumptions that are not appropriate for p-version finite elements. Two methods for determining internal radiation heat transfer are developed for one and two-dimensional p-version finite elements. In the first method, higher-order elements are divided into a number of sub-elements. Traditional methods are used to determine radiation heat flux along each sub-element and then mapped back to the parent element. In the second method, the radiation heat transfer equations are numerically integrated over the higher-order element. Comparisons with analytical solutions show that the integration scheme is generally more accurate than the sub-element method. Comparison to results from traditional finite elements shows that significant reduction in the number of elements in the mesh is possible using higher-order (p-version) finite elements.

  14. HEAT AND WATER TRANSPORT IN A POLYMER ELECTROLYTE FUEL CELL

    SciTech Connect

    Mukherjee, Partha P

    2010-01-01

    In the present scenario of a global initiative toward a sustainable energy future, the polymer electrolyte fuel cell (PEFC) has emerged as one of the most promising alternative energy conversion devices for various applications. Despite tremendous progress in recent years, a pivotal performance limitation in the PEFC comes from liquid water transport and the resulting flooding phenomena. Liquid water blocks the open pore space in the electrode and the fibrous diffusion layer leading to hindered oxygen transport. The electrode is also the only component in the entire PEFC sandwich which produces waste heat from the electrochemical reaction. The cathode electrode, being the host to several competing transport mechanisms, plays a crucial role in the overall PEFC performance limitation. In this work, an electrode model is presented in order to elucidate the coupled heat and water transport mechanisms. Two scenarios are specifically considered: (1) conventional, Nafion impregnated, three-phase electrode with the hydrated polymeric membrane phase as the conveyer of protons where local electro-neutrality prevails; and (2) ultra-thin, two-phase, nano-structured electrode without the presence of ionomeric phase where charge accumulation due to electro-statics in the vicinity of the membrane-CL interface becomes important. The electrode model includes a physical description of heat and water balance along with electrochemical performance analysis in order to study the influence of electro-statics/electro-migration and phase change on the PEFC electrode performance.

  15. Experimental study of heat transfer in a 7-element bundle cooled with supercritical Freon-12

    SciTech Connect

    Richards, G.; Shelegov, A. S.; Kirillov, P. L.; Pioro, I. L.; Harvel, G.

    2012-07-01

    Experimental data on Supercritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are technical difficulties in testing and experimental costs at high pressures, temperatures and heat fluxes. Also, there are only a few SCW experimental setups currently in the world capable of providing data. Supercritical Water-cooled nuclear Reactors (SCWRs), as one of the six concepts of Generation IV reactors, cannot be designed without such data. Therefore, a preliminary approach uses modeling fluids such as carbon dioxide and refrigerants instead of water is practical. In particularly, experiments in supercritical refrigerant-cooled bundles can be used. One of the SC modeling fluids typically used is Freon-12 (R-12) with the critical pressure of 4.136 MPa and the critical temperature of 111.97 deg. C. These conditions correspond to the critical pressure of 22.064 MPa and critical temperature of 373.95 deg. C in water. A set of experimental data obtained at the Inst. of Physics and Power Engineering (IPPE, Obninsk, Russia) in a vertically-oriented bundle cooled with supercritical R-12 was analyzed. This dataset consisted of 20 runs. The test section was 7-element bundle installed in a hexagonal flow channel with 3 grid spacers. Data was collected at pressures of approximately 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudo-critical temperature. The values of mass flux were ranged from 400 to 1320 kg/m{sup 2}s and inlet temperatures ranged from 72 to 120 deg. C. The test section consisted of fuel-element simulators that were 9.5 mm in OD with the total heated length of about 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 different thermocouples. Analysis of the data has confirmed that there are three distinct heat-transfer regimes for forced convention in supercritical fluids: 1) Normal heat transfer; 2) Deteriorated heat

  16. 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.

  17. An environmentally benign soybean derived fuel as a blending stock or replacement for home heating oil.

    PubMed

    Mushrush, G; Beal, E J; Spencer, G; Wynne, J H; Lloyd, C L; Hughes, J M; Walls, C L; Hardy, D R

    2001-05-01

    The use of bio-derived materials both as fuels and/or as blending stocks becomes more attractive as the price of middle distillate fuels, especially home heating oil, continues to rise. Historically, many biomass and agricultural derived materials have been suggested. One of the most difficult problems encountered with home heating oil is that of storage stability. High maintenance costs associated with home heating oil are, in large part, because of this stability problem. In the present research, Soygold, a soybean derived fuel, was added in concentrations of 10%-20% to both a stable middle distillate fuel and an unstable home heating oil. Fuel instability in this article will be further related to the organo-nitrogen compounds present. The soy-fuel mixtures proved stable, and the addition of the soy liquid enhanced both the combustion properties, and dramatically improved the stability of the unstable home heating oil. PMID:11460320

  18. 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.

  19. Finite element analysis of monolithic solid oxide fuel cells

    SciTech Connect

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

    1992-01-01

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

  20. Finite element analysis of monolithic solid oxide fuel cells

    SciTech Connect

    Saigal, A.; Majumdar, S.

    1992-04-01

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

  1. Heat shock stimulation of a tilapia heat shock protein 70 promoter is mediated by a distal element.

    PubMed Central

    Molina, A; Di Martino, E; Martial, J A; Muller, M

    2001-01-01

    We reported previously that a tilapia (Oreochromis mossambicus) heat shock protein 70 (HSP70) promoter is able to confer heat shock response on a reporter gene after transient expression both in cell culture and in microinjected zebrafish embryos. Here we present the first functional analysis of a fish HSP70 promoter, the tiHSP70 promoter. Using transient expression experiments in carp EPC (epithelioma papulosum cyprini) cells and in microinjected zebrafish embryos, we show that a distal heat shock response element (HSE1) at approx. -800 is predominantly responsible for the heat shock response of the tiHSP70 promoter. This element specifically binds an inducible transcription factor, most probably heat shock factor, and a constitutive factor. The constitutive complex is not observed with the non-functional, proximal HSE3 sequence, suggesting that both factors are required for the heat shock response mediated by HSE1. PMID:11368761

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

    SciTech Connect

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

    1982-07-01

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

  3. 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.

  4. 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)

  5. Environmental assessment for radioisotope heat source fuel processing and fabrication

    SciTech Connect

    Not Available

    1991-07-01

    DOE has prepared an Environmental Assessment (EA) for radioisotope heat source fuel processing and fabrication involving existing facilities at the Savannah River Site (SRS) near Aiken, South Carolina and the Los Alamos National Laboratory (LANL) near Los Alamos, New Mexico. The proposed action is needed to provide Radioisotope Thermoelectric Generators (RTG) to support the National Aeronautics and Space Administration's (NASA) CRAF and Cassini Missions. Based on the analysis in the EA, DOE has determined that the proposed action does not constitute a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, an Environmental Impact Statement is not required. 30 refs., 5 figs.

  6. Utilization of waste heat in trucks for increased fuel economy

    NASA Technical Reports Server (NTRS)

    Leising, C. J.; Purohit, G. P.; Degrey, S. P.; Finegold, J. G.

    1978-01-01

    The waste heat utilization concepts include preheating, regeneration, turbocharging, turbocompounding, and Rankine engine compounding. Predictions are based on fuel-air cycle analyses, computer simulation, and engine test data. All options are evaluated in terms of maximum theoretical improvements, but the Diesel and adiabatic Diesel are also compared on the basis of maximum expected improvement and expected improvement over a driving cycle. The study indicates that Diesels should be turbocharged and aftercooled to the maximum possible level. The results reveal that Diesel driving cycle performance can be increased by 20% through increased turbocharging, turbocompounding, and Rankine engine compounding. The Rankine engine compounding provides about three times as much improvement as turbocompounding but also costs about three times as much. Performance for either can be approximately doubled if applied to an adiabatic Diesel.

  7. Experimental Investigation of Turbine Vane Heat Transfer for Alternative Fuels

    SciTech Connect

    Nix, Andrew Carl

    2015-03-23

    The focus of this program was to experimentally investigate advanced gas turbine cooling schemes and the effects of and factors that contribute to surface deposition from particulate matter found in coal syngas exhaust flows on turbine airfoil heat transfer and film cooling, as well as to characterize surface roughness and determine the effects of surface deposition on turbine components. The program was a comprehensive, multi-disciplinary collaborative effort between aero-thermal and materials faculty researchers and the Department of Energy, National Energy Technology Laboratory (NETL). The primary technical objectives of the program were to evaluate the effects of combustion of syngas fuels on heat transfer to turbine vanes and blades in land-based power generation gas turbine engines. The primary questions to be answered by this investigation were; What are the factors that contribute to particulate deposition on film cooled gas turbine components? An experimental program was performed in a high-temperature and pressure combustion rig at the DOE NETL; What is the effect of coal syngas combustion and surface deposition on turbine airfoil film cooling? Deposition of particulate matter from the combustion gases can block film cooling holes, decreasing the flow of the film coolant and the film cooling effectiveness; How does surface deposition from coal syngas combustion affect turbine surface roughness? Increased surface roughness can increase aerodynamic losses and result in decreased turbine hot section efficiency, increasing engine fuel consumption to maintain desired power output. Convective heat transfer is also greatly affected by the surface roughness of the airfoil surface; Is there any significant effect of surface deposition or erosion on integrity of turbine airfoil thermal barrier coatings (TBC) and do surface deposits react with the TBC in any way to decrease its thermal insulating capability? Spallation and erosion of TBC is a persistent problem in

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

    SciTech Connect

    Hayes, S.L.

    1993-12-01

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

  9. Maintenance and storage of fuel oil for residential heating systems: A guide for residential heating system maintenance personnel

    SciTech Connect

    Litzke, Wai-Lin

    1992-12-01

    The quality of No. 2 fuel affects the performance of the heating system and is an important parameter in the proper and efficient operation of an oil-burning system. The physical and chemical characteristics of the fuel can affect the flow, atomization and combustion processes, all of which help to define and limit the overall performance of the heating system. The use of chemical additives by fuel oil marketershas become more common as a method of improving the quality of the fuel, especially for handling and storage. Numerous types of additives are available, but reliable information on their effectiveness and proper use is limited. This makes selecting an additive difficult in many situations. Common types of problems that contribute to poor fuel quality and how they affect residential heating equipment are identified inof this booklet. It covers the key items that are needed in an effective fuel quality monitoring program, such as what to look for when evaluating the quality of fuel as it is received from a supplier, or how to assess fuel problems associated with poor storage conditions. References to standard procedures and brief descriptions of the procedures also are given. Approaches for correcting a fuel-related problem, including the potential uses of chemical additives are discussed. Different types of additives are described to help users understand the functions and limitations of chemical treatment. Tips on how to select andeffectively use additives also are included. Finally, the importance of preventative maintenance in any fuel monitoring program is emphasized.

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

    SciTech Connect

    Sholtis, Joseph A. Jr.

    1982-07-01

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

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

    PubMed

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

    1985-01-01

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

  12. Experimental Study of Fuel Heating at Low Temperatures in a Wing Tank Model, Volume 1

    NASA Technical Reports Server (NTRS)

    Stockemer, F. J.

    1981-01-01

    Scale model fuel heating systems for use with aviation hydrocarbon fuel at low temperatures were investigated. The effectiveness of the heating systems in providing flowability and pumpability at extreme low temperature when some freezing of the fuel would otherwise occur is evaluated. The test tank simulated a section of an outer wing tank, and was chilled on the upper and lower surfaces. Turbine engine lubricating oil was heated, and recirculating fuel transferred the heat. Fuels included: a commercial Jet A; an intermediate freeze point distillate; a higher freeze point distillate blended according to Experimental Referee Broadened Specification guidelines; and a higher freeze point paraffinic distillate used in a preceding investigation. Each fuel was chilled to selected temperature to evaluate unpumpable solid formation (holdup). Tests simulating extreme cold weather flight, without heating, provided baseline fuel holdup data. Heating and recirculating fuel increased bulk temperature significantly; it had a relatively small effect on temperature near the bottom of the tank. Methods which increased penetration of heated fuel into the lower boundary layer improved the capability for reducing holdup.

  13. 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

  14. Electrical heating tests of uranium dioxide external fuel configuration at emitter temperature of 1900 K

    NASA Technical Reports Server (NTRS)

    Diianni, D. C.; Mayer, J. T.

    1974-01-01

    Testing of two fuel clad specimens for thermionic reactor application is described. The annular UO2 fuel was clad on both sides with tungsten; heat rejection was radially inward. The tests were intended to study inner clad stability, fuel redistribution, and fuel melting problems. The specimens were tested in a vacuum chamber using electron bombardment heating. Fuel structural changes were studied using periodic gammagraphs and posttest metallography. The first specimen test was terminated at 50 hours because of a braze failure. The second specimen was tested for 240 hours when an outer clad leak developed due to a tungsten-water reaction. The fuel developed numerous cracks on cooldown but the inner clad remained dimensionally stable. The fuel cover gas did not impede the rate of fuel redistribution. Posttest examination showed the fuel had not melted during operation.

  15. 46 CFR 147.50 - Fuel for cooking, heating, and lighting.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Fuel for cooking, heating, and lighting. 147.50 Section 147.50 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) DANGEROUS CARGOES HAZARDOUS SHIPS' STORES Stowage and Other Special Requirements for Particular Materials § 147.50 Fuel for cooking, heating, and lighting. (a) Flammable...

  16. 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.

  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. Conceptual study of measures against heat generation for TRU fuel fabrication system

    SciTech Connect

    Kawaguchi, Koichi; Namekawa, Takashi

    2007-07-01

    To lower the reprocessing cost and the environmental burden, the Japan Atomic Energy Agency (JAEA) has developed low decontamination TRU fuel fabrication system. TRU fuel contains MA of 1.2 to 5 wt% and its decay heat is estimated a few tens W/kg-HM. As the heat affects fuel quality through oxidation of fuel material and members, it is necessary to remove decay heat. In this work, authors designed concepts of the measures against heat generation at typical equipments using with the thermal hydraulics analysis technique. As a result, it is shown that it is possible to cool fuel materials with specific heat generation up to 20 W/kg-HM enough, though more detailed study is required for comprehensive equipments. (authors)

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

    NASA Technical Reports Server (NTRS)

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

    1974-01-01

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

  20. Design and evaluation of aircraft heat source systems for use with high-freezing point fuels

    NASA Technical Reports Server (NTRS)

    Pasion, A. J.

    1979-01-01

    The objectives were the design, performance and economic analyses of practical aircraft fuel heating systems that would permit the use of high freezing-point fuels on long-range aircraft. Two hypothetical hydrocarbon fuels with freezing points of -29 C and -18 C were used to represent the variation from current day jet fuels. A Boeing 747-200 with JT9D-7/7A engines was used as the baseline aircraft. A 9300 Km mission was used as the mission length from which the heat requirements to maintain the fuel above its freezing point was based.

  1. Three isoparametric solid elements for NASTRAN. [for static, dynamic, buckling, and heat transfer analyses

    NASA Technical Reports Server (NTRS)

    Johnson, S. E.; Field, E. I.

    1973-01-01

    Linear, quadratic, and cubic isoparametric hexahedral solid elements have been added to the element library of NASTRAN. These elements are available for static, dynamic, buckling, and heat-transfer analyses. Because the isoparametric element matrices are generated by direct numerical integration over the volume of the element, variations in material properties, temperatures, and stresses within the elements are represented in the computations. In order to compare the accuracy of the new elements, three similar models of a slender cantilever were developed, one for each element. All elements performed well. As expected, however, the linear element model yielded excellent results only when shear behavior predominated. In contrast, the results obtained from the quadratic and cubic element models were excellent in both shear and bending.

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

    SciTech Connect

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

    2008-07-15

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

  3. Near-field heat transfer at the spent fuel test-climax: a comparison of measurements and calculations

    SciTech Connect

    Patrick, W.C.; Montan, D.N.; Ballou, L.B.

    1981-08-21

    The Spent Fuel Test in the Climax granitic stock at the DOE Nevada Test Site is a test of the feasibility of storage and retrieval of spent nuclear reactor fuel in a deep geologic environment. Eleven spent fuel elements, together with six thermally identical electrical resistance heaters and 20 peripheral guard heaters, are emplaced 420 m below surface in a three-drift test array. This array was designed to simulate the near-field effects of thousands of canisters of nuclear waste and to evaluate the effects of heat alone, and heat plus ionizing radiation on the rock. Thermal calculations and measurements are conducted to determine thermal transport from the spent fuel and electrical resistance heaters. Calculations associated with the as-built Spent Fuel Test geometry and thermal source histories are presented and compared with thermocouple measurements made throughout the test array. Comparisons in space begin at the spent fuel canister and include the first few metres outside the test array. Comparisons in time begin at emplacement and progress through the first year of thermal loading in this multi-year test.

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

    SciTech Connect

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

    1988-07-01

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

  5. TLD measurements of gamma heating in heavy elements.

    NASA Technical Reports Server (NTRS)

    Reilly, H. J.; Robinson, R. A.; Peters, L. E., Jr.

    1971-01-01

    Measurements and calculations of gamma heating in polyethylene and lead containers were done and compared. The objective was to provide a workable method of getting good values for gamma heating in in-pile experiments containing materials of high atomic numbers. It was inferred that a combination of thermoluminescent dosimeter measurements, using Bragg-Gray theory, with photon transport calculations using the ANISN computer program, would meet this objective.

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

    DOEpatents

    Erwin, J.H.

    1962-12-11

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

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

    SciTech Connect

    Pitner, A.L.

    1998-09-11

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

  8. Carbon or graphite foam as a heating element and system thereof

    DOEpatents

    Ott, Ronald D [Knoxville, TN; McMillan, April D [Knoxville, TN; Choudhury, Ashok [Oak Ridge, TN

    2004-05-04

    A temperature regulator includes at least one electrically conductive carbon foam element. The foam element includes at least two locations adapted for receiving electrical connectors thereto for heating a fluid, such as engine oil. A combustion engine includes an engine block and at least one carbon foam element, the foam element extending into the engine block or disposed in thermal contact with at least one engine fluid.

  9. Evaluation of a Passive Heat Exchanger Based Cooling System for Fuel Cell Applications

    NASA Technical Reports Server (NTRS)

    Colozza, Anthony J.; Burke, Kenneth A.

    2011-01-01

    Fuel cell cooling is conventionally performed with an actively controlled, dedicated coolant loop that exchanges heat with a separate external cooling loop. To simplify this system the concept of directly cooling a fuel cell utilizing a coolant loop with a regenerative heat exchanger to preheat the coolant entering the fuel cell with the coolant exiting the fuel cell was analyzed. The preheating is necessary to minimize the temperature difference across the fuel cell stack. This type of coolant system would minimize the controls needed on the coolant loop and provide a mostly passive means of cooling the fuel cell. The results indicate that an operating temperature of near or greater than 70 C is achievable with a heat exchanger effectiveness of around 90 percent. Of the heat exchanger types evaluated with the same type of fluid on the hot and cold side, a counter flow type heat exchanger would be required which has the possibility of achieving the required effectiveness. The number of heat transfer units required by the heat exchanger would be around 9 or greater. Although the analysis indicates the concept is feasible, the heat exchanger design would need to be developed and optimized for a specific fuel cell operation in order to achieve the high effectiveness value required.

  10. Micro-tubular flame-assisted fuel cells for micro-combined heat and power systems

    NASA Astrophysics Data System (ADS)

    Milcarek, Ryan J.; Wang, Kang; Falkenstein-Smith, Ryan L.; Ahn, Jeongmin

    2016-02-01

    Currently the role of fuel cells in future power generation is being examined, tested and discussed. However, implementing systems is more difficult because of sealing challenges, slow start-up and complex thermal management and fuel processing. A novel furnace system with a flame-assisted fuel cell is proposed that combines the thermal management and fuel processing systems by utilizing fuel-rich combustion. In addition, the flame-assisted fuel cell furnace is a micro-combined heat and power system, which can produce electricity for homes or businesses, providing resilience during power disruption while still providing heat. A micro-tubular solid oxide fuel cell achieves a significant performance of 430 mW cm-2 operating in a model fuel-rich exhaust stream.

  11. Heat recovery subsystem and overall system integration of fuel cell on-site integrated energy systems

    NASA Technical Reports Server (NTRS)

    Mougin, L. J.

    1983-01-01

    The best HVAC (heating, ventilating and air conditioning) subsystem to interface with the Engelhard fuel cell system for application in commercial buildings was determined. To accomplish this objective, the effects of several system and site specific parameters on the economic feasibility of fuel cell/HVAC systems were investigated. An energy flow diagram of a fuel cell/HVAC system is shown. The fuel cell system provides electricity for an electric water chiller and for domestic electric needs. Supplemental electricity is purchased from the utility if needed. An excess of electricity generated by the fuel cell system can be sold to the utility. The fuel cell system also provides thermal energy which can be used for absorption cooling, space heating and domestic hot water. Thermal storage can be incorporated into the system. Thermal energy is also provided by an auxiliary boiler if needed to supplement the fuel cell system output. Fuel cell/HVAC systems were analyzed with the TRACE computer program.

  12. 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.

  13. Process for massively hydriding zirconium--uranium fuel elements

    DOEpatents

    Katz, N.H.

    1973-12-01

    A method is described of hydriding uranium-zirconium alloy by heating the alloy in a vacuum, introducing hydrogen and maintaining an elevated temperature until occurrence of the beta--delta phase transformation and isobarically cooling the composition. (Official Gazette)

  14. Synergistic Effects of Toxic Elements on Heat Shock Proteins

    PubMed Central

    Mahmood, Khalid; Mahmood, Qaisar; Irshad, Muhammad; Hussain, Jamshaid

    2014-01-01

    Heat shock proteins show remarkable variations in their expression levels under a variety of toxic conditions. A research span expanded over five decades has revealed their molecular characterization, gene regulation, expression patterns, vast similarity in diverse groups, and broad range of functional capabilities. Their functions include protection and tolerance against cytotoxic conditions through their molecular chaperoning activity, maintaining cytoskeleton stability, and assisting in cell signaling. However, their role as biomarkers for monitoring the environmental risk assessment is controversial due to a number of conflicting, validating, and nonvalidating reports. The current knowledge regarding the interpretation of HSPs expression levels has been discussed in the present review. The candidature of heat shock proteins as biomarkers of toxicity is thus far unreliable due to synergistic effects of toxicants and other environmental factors. The adoption of heat shock proteins as “suit of biomarkers in a set of organisms” requires further investigation. PMID:25136596

  15. 46 CFR 147.50 - Fuel for cooking, heating, and lighting.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 5 2012-10-01 2012-10-01 false Fuel for cooking, heating, and lighting. 147.50 Section..., heating, and lighting. (a) Flammable and combustible liquids and gases not listed in this section are prohibited for cooking, heating, or lighting on any vessel, with the exception of combustible liquids...

  16. 46 CFR 147.50 - Fuel for cooking, heating, and lighting.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 5 2011-10-01 2011-10-01 false Fuel for cooking, heating, and lighting. 147.50 Section..., heating, and lighting. (a) Flammable and combustible liquids and gases not listed in this section are prohibited for cooking, heating, or lighting on any vessel, with the exception of combustible liquids...

  17. 46 CFR 147.50 - Fuel for cooking, heating, and lighting.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 5 2014-10-01 2014-10-01 false Fuel for cooking, heating, and lighting. 147.50 Section..., heating, and lighting. (a) Flammable and combustible liquids and gases not listed in this section are prohibited for cooking, heating, or lighting on any vessel, with the exception of combustible liquids...

  18. 46 CFR 147.50 - Fuel for cooking, heating, and lighting.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 5 2013-10-01 2013-10-01 false Fuel for cooking, heating, and lighting. 147.50 Section..., heating, and lighting. (a) Flammable and combustible liquids and gases not listed in this section are prohibited for cooking, heating, or lighting on any vessel, with the exception of combustible liquids...

  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. Effect of fuel density and heating value on ram-jet airplane range

    NASA Technical Reports Server (NTRS)

    Henneberry, Hugh M

    1952-01-01

    An analytical investigation of the effects of fuel density and heating value on the cruising range of a ram-jet airplane was made. Results indicate that with present-day knowledge of chemical fuels, neither very high nor very low fuel densities have any advantages for long-range flight. Of the fuels investigated, the borohydrides and metallic boron have the greatest range potential. Aluminum and aluminum hydrocarbon slurries were inferior to pure hydrocarbon fuel and boron-hydrocarbon slurries were superior on a range basis. It was concluded that the practical difficulties associated with the use of liquid hydrogen fuel cannot be justified on a range basis.

  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. Nondestructive inspection of General Purpose Heat Source (GPHS) fueled clad girth welds

    SciTech Connect

    Reimus, M.A.; George, T.G.; Lynch, C.; Padilla, M.; Moniz, P.; Guerrero, A.; Moyer, M.W.; Placr, A.

    1998-01-01

    The General-Purpose Heat Source (GPHS) provides power for space missions by transmitting the heat of {sup 238}Pu decay to an array of thermoelectric elements. The GPHS is fabricated using an iridium-alloy to contain the {sup 238}PuO{sub 2} fuel pellet. GPHS capsules will be utilized in the upcoming Cassini mission to explore Saturn and its moons. The physical integrity of the girth weld is important to mission safety and performance. Because past experience had revealed a potential for initiation of small cracks in the girth weld overlap zone, a nondestructive inspection of each capsule weld is required. An ultrasonic method was used to inspect the welds of capsules fabricated for the Galileo mission. The instrument, transducer, and method used were state of the art at the time (early 1980s). The ultrasonic instrumentation and methods used to inspect the Cassini GPHSs was significantly upgraded from those used for the Galileo mission. GPHSs that had ultrasonic reflectors in excess of the reject specification level were subsequently inspected with radiography to provide additional engineering data used to accept/reject the heat source. This paper describes the Galileo-era ultrasonic instrumentation and methods and the subsequent upgrades made to support testing of Cassini GPHSs. Also discussed is the data obtained from radiographic examination and correlation to ultrasonic examination results. {copyright} {ital 1998 American Institute of Physics.}

  3. Nondestructive inspection of General Purpose Heat Source (GPHS) fueled clad girth welds

    SciTech Connect

    Reimus, M. A. H.; George, T. G.; Lynch, C.; Padilla, M.; Moniz, P.; Guerrero, A.; Moyer, M. W.; Placr, A.

    1998-01-15

    The General-Purpose Heat Source (GPHS) provides power for space missions by transmitting the heat of {sup 238}Pu decay to an array of thermoelectric elements. The GPHS is fabricated using an iridium-alloy to contain the {sup 238}PuO{sub 2} fuel pellet. GPHS capsules will be utilized in the upcoming Cassini mission to explore Saturn and its moons. The physical integrity of the girth weld is important to mission safety and performance. Because past experience had revealed a potential for initiation of small cracks in the girth weld overlap zone, a nondestructive inspection of each capsule weld is required. An ultrasonic method was used to inspect the welds of capsules fabricated for the Galileo mission. The instrument, transducer, and method used were state of the art at the time (early 1980s). The ultrasonic instrumentation and methods used to inspect the Cassini GPHSs was significantly upgraded from those used for the Galileo mission. GPHSs that had ultrasonic reflectors in excess of the reject specification level were subsequently inspected with radiography to provide additional engineering data used to accept/reject the heat source. This paper describes the Galileo-era ultrasonic instrumentation and methods and the subsequent upgrades made to support testing of Cassini GPHSs. Also discussed is the data obtained from radiographic examination and correlation to ultrasonic examination results.

  4. Boundary element techniques - Applications in stress analysis and heat transfer

    SciTech Connect

    Brebbia, C.A.; Venturini, W.S.

    1987-01-01

    This volume includes contributions in the field of stress analysis, soil and rock mechanics, non-linear problems, dynamics and vibrations, plate bending and heat transfer. The companion volume includes contributions dealing with viscous and inviscid fluid flow, aerodynamics and hydrodynamics applications, elastostatics and computational and mathematical aspects.

  5. 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.

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

    SciTech Connect

    Maassen, D.P., Fluor Daniel Hanford

    1997-03-11

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

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

    SciTech Connect

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

    1999-09-24

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

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

    DOEpatents

    Shore, Lawrence; Matlin, Ramail; Heinz, Robert

    2012-06-26

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

  9. The Northeast heating fuel market: Assessment and options

    SciTech Connect

    2000-07-01

    In response to a Presidential request, this study examines how the distillate fuel oil market (and related energy markets) in the Northeast behaved in the winter of 1999-2000, explains the role played by residential, commercial, industrial, and electricity generation sector consumers in distillate fuel oil markets and describes how that role is influenced by the structure of tie energy markets in the Northeast. In addition, this report explores the potential for nonresidential users to move away from distillate fuel oil and how this might impact future prices, and discusses conversion of distillate fuel oil users to other fuels over the next 5 years. Because the President's and Secretary's request focused on converting factories and other large-volume users of mostly high-sulfur distillate fuel oil to other fuels, transportation sector use of low-sulfur distillate fuel oil is not examined here.

  10. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    SciTech Connect

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

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

    SciTech Connect

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

    1995-12-31

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

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

    SciTech Connect

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

    1980-12-01

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

  13. Molten tin reprocessing of spent nuclear fuel elements

    DOEpatents

    Heckman, Richard A.

    1983-01-01

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

  14. Analyzing the possibility of constructing the air heating system for an integrated solid fuel gasification combined-cycle power plant

    NASA Astrophysics Data System (ADS)

    Mikula, V. A.; Ryzhkov, A. F.; Val'tsev, N. V.

    2015-11-01

    Combined-cycle power plants operating on solid fuel have presently been implemented only in demonstration projects. One of possible ways for improving such plants consists in making a shift to hybrid process circuits of integrated gasification combined-cycle plants with external firing of solid fuel. A high-temperature air heater serving to heat compressed air is a key element of the hybrid process circuit. The article describes application of a high-temperature recuperative metal air heater in the process circuit of an integrated gasification combined-cycle power plant (IGCC). The available experience with high-temperature air heating is considered, and possible air heater layout arrangements are analyzed along with domestically produced heat-resistant grades of steel suitable for manufacturing such air heater. An alternative (with respect to the traditional one) design is proposed, according to which solid fuel is fired in a noncooled furnace extension, followed by mixing the combustion products with recirculation gases, after which the mixture is fed to a convective air heater. The use of this design makes it possible to achieve considerably smaller capital outlays and operating costs. The data obtained from thermal and aerodynamic calculations of the high-temperature air heater with a thermal capacity of 258 MW for heating air to a temperature of up to 800°C for being used in the hybrid process circuit of a combined-cycle power plant are presented.

  15. ORGANIC COMBUSTION FINGERPRINTS OF THREE COMMON HOME HEATING FUELS

    EPA Science Inventory

    The paper discusses the chemical structures of three common home eating fuels: wood, coal, and No. 2 fuel oil. GC and GC/MS data are then presented which demonstrate how the thermal destruction of each fuel results in the production of a characteristic group of organic "fingerpri...

  16. Optimal Heat Collection Element Shapes for Parabolic Trough Concentrators

    SciTech Connect

    Bennett, C

    2007-11-15

    For nearly 150 years, the cross section of the heat collection tubes used at the focus of parabolic trough solar concentrators has been circular. This type of tube is obviously simple and easily fabricated, but it is not optimal. It is shown in this article that the optimal shape, assuming a perfect parabolic figure for the concentrating mirror, is instead oblong, and is approximately given by a pair of facing parabolic segments.

  17. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-12-31

    This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

  18. Sweat mineral element responses during 7 h of exercise-heat stress

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Uncertainty exists regarding the effect of sustained sweating on sweat mineral element composition. This study determined the effect of multiple hours of exercise-heat stress on sweat mineral concentrations. Seven heat acclimated subjects (6 males, 1 female) completed 5 consecutive 60 min bouts of...

  19. 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.

  20. 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.

  1. 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.

  2. 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.

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

    SciTech Connect

    D. D. Keiser; J. I. Cole

    2007-09-01

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

  4. Experimental and numerical simulations of flow and heat transfer in heat exchanger elements using liquid crystal thermography

    NASA Astrophysics Data System (ADS)

    Stasiek, Jan; Ciofalo, Michele; Wierzbowski, Maciej

    2004-05-01

    Experimental and numerical investigation of heat transfer and fluid flow were conducted for classic heat exchanger elements (flat plate with fin-tubes in-line, staggered and with vortex generators) and corrugated-undulated ducts under transitional and weakly turbulent conditions. The dependence of average heat transfer and pressure drop on Reynolds number and geometrical parameters was investigated. Distributions of local heat transfer coefficient were obtained by using liquid crystal thermography and surface-averaged values were computed. Three-dimensional numerical simulations were conducted by a finite-volume method using a low-Reynolds number k-ɛ model under the assumption of fully developed flow. Computed flow fields provided otherwise inaccessible information on the flow patterns and the mechanisms of heat transfer enhancement.

  5. An Investigation of Size-Dependent Concentration of Trace Elements in Aerosols Emitted from the Oil-Fired Heating Plants

    NASA Technical Reports Server (NTRS)

    Singh, J. J.; Sentell, R. J.; Khandelwal, G. S.

    1976-01-01

    Aerosols emitted from two oil-fired heating plants were aerodynamically separated into eight size groups and were analyzed using the photon-induced X-ray emission (PIXE) technique. It was found that Zn, Mo, Ag, and Pb, and (to a lesser extent) Cd, have a tendency to concentrate preferentially on the smaller aerosols. All of these elements, in certain chemical forms, are known to be toxic. Zinc and molybdenum, although present in low concentrations in the parent fuels, show the strongest tendencies to be concentrated in finer aerosols. Selenium, previously reported to show a very strong tendency to concentration in finer fly ash from coal-fired power plants shows little preference for surface residence. Vanadium, which occurs in significant concentration in the oil fuels for both plants, also shows little preference for surface concentration. Even though the absolute concentrations of the toxic elements involved are well below the safety levels established by the National Institute for Occupational Safety and Health (NIOSH), it would be advisable to raise the heights of the heating-plant exhaust chimneys well above the neighborhood buildings to insure more efficient aerosol dispersal.

  6. Configuring a fuel cell based residential combined heat and power system

    NASA Astrophysics Data System (ADS)

    Ahmed, Shabbir; Papadias, Dionissios D.; Ahluwalia, Rajesh K.

    2013-11-01

    The design and performance of a fuel cell based residential combined heat and power (CHP) system operating on natural gas has been analyzed. The natural gas is first converted to a hydrogen-rich reformate in a steam reformer based fuel processor, and the hydrogen is then electrochemically oxidized in a low temperature polymer electrolyte fuel cell to generate electric power. The heat generated in the fuel cell and the available heat in the exhaust gas is recovered to meet residential needs for hot water and space heating. Two fuel processor configurations have been studied. One of the configurations was explored to quantify the effects of design and operating parameters, which include pressure, temperature, and steam-to-carbon ratio in the fuel processor, and fuel utilization in the fuel cell. The second configuration applied the lessons from the study of the first configuration to increase the CHP efficiency. Results from the two configurations allow a quantitative comparison of the design alternatives. The analyses showed that these systems can operate at electrical efficiencies of ∼46% and combined heat and power efficiencies of ∼90%.

  7. Major and minor element site occupancies in heated natural forsterite

    SciTech Connect

    Smyth, J.R.; Taftoe, J.

    1982-09-01

    Using a new analytical transmission electron microscopic technique known as CHannelling Enhanced X-ray Emission (CHEXE) spectroscopy, the M-site occupancies of Fe, Ni, Mn, and Ca have been determined in a natural forsteritic olivine (Fo/sub 91/) heat treated at different temperatures. The sample was taken as a single olivine grain from a spinel peridotite inclusion in an alkali basalt and contains 0.36 wt% NiO, 0.07 wt% MnO, and 0.09 wt% CaO. In the non-heat-treated sample, 49.6 at % of the Fe, 97 +/- 5 at % of the Mn in the sample occupy the M1 site. In the present study of samples quenched from different temperatures, the fraction of the Ni present at M1 is 87 +/- 5% (6 days at 300/sup 0/C), 83 +/- 5% (48h at 600/sup 0/C), 83 +/- 5% (45h at 900/sup 0/C) and 80 +/- 5% (24h at 1050/sup 0/C). The authors observed a lesser tendency for Ni to order than postulated by previous workers for Ni-rich olivines. For Mn, typically 15% of the atoms occupy M1 in the heat treated samples. No significant deviation from complete ordering into M2 was observed for Ca. The Fe atoms are completely disordered with 50 +/- 1% at each M-site, except for a weak deviation at 300/sup 0/C with 47.1 +/- 1% at M1. The study indicates that exchange of cations between M-sites may begin as low as 300/sup 0/C. This implies that Ni and Mn distribution in natural olivines may be a useful indicator of cooling rate in rapidly cooled rocks.

  8. Finite element formulation for transient heat treat problems

    NASA Technical Reports Server (NTRS)

    Mullen, R. L.; Hendricks, R. C.

    1983-01-01

    The macrothermomechanical behavior of materials subjected to rapid thermal or mechanical loading such as occurs in most heat treatments is described. The equations are developed for Lagrangian, Eulerian, and intermediary kinematic descriptions and are independent of the constitutive laws and the equation of state; they can be solved numerically for a specified material and boundary conditions. The coupled transport effects between dissipation and energy are included. The conventional linearized stability approach indicates the numerical procedure to be stable, with certain restriction on the time step size.

  9. Install Waste Heat Recovery Systems for Fuel-Fired Furnaces (English/Chinese) (Fact Sheet)

    SciTech Connect

    Not Available

    2011-10-01

    Chinese translation of ITP fact sheet about installing Waste Heat Recovery Systems for Fuel-Fired Furnaces. For most fuel-fired heating equipment, a large amount of the heat supplied is wasted as exhaust or flue gases. In furnaces, air and fuel are mixed and burned to generate heat, some of which is transferred to the heating device and its load. When the heat transfer reaches its practical limit, the spent combustion gases are removed from the furnace via a flue or stack. At this point, these gases still hold considerable thermal energy. In many systems, this is the greatest single heat loss. The energy efficiency can often be increased by using waste heat gas recovery systems to capture and use some of the energy in the flue gas. For natural gas-based systems, the amount of heat contained in the flue gases as a percentage of the heat input in a heating system can be estimated by using Figure 1. Exhaust gas loss or waste heat depends on flue gas temperature and its mass flow, or in practical terms, excess air resulting from combustion air supply and air leakage into the furnace. The excess air can be estimated by measuring oxygen percentage in the flue gases.

  10. Theoretical Rationale of Heating Block for Testing Bench of Aerospace Crafts Thermal Protection Elements

    NASA Astrophysics Data System (ADS)

    Petrova, Anna A.; Reznik, Sergey V.

    2016-02-01

    The theoretical rationale for the structural layout of a testing bench with zirconium dioxide heating elements on the basis of modelling radiative-conductive heat transfer are presented. The numerical simulation of radiative-conductive heat transfer for the two-dimensional scaled model of the testing segment with the finite-element analysis software package Ansys 15.0 are performed. The simulation results showed that for the selected layout of the heaters the temperature non-uniformity along the length of the sample over time will not exceed 3 % even at a temperature of 2000 K.

  11. 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.

  12. 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.

  13. 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.

  14. 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/.

  15. 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].

  16. Lunar surface heat flow mapping from radioactive elements measured by Lunar Prospector

    NASA Astrophysics Data System (ADS)

    Zhang, Dan; Li, Xiongyao; Li, Qingxia; Lang, Liang; Zheng, Yongchun

    2014-06-01

    An accurate estimate of global surface heat flow is important because it provides strong constraints on interior thermal model and understanding of the thermal state and geologic evolution of the Moon. In this paper, a distribution map of lunar surface heat flow is derived from calibrated Lunar Prospector gamma-ray spectrometer data (K, U and Th abundances). It shows that surface heat flow varies regionally from about 10.6 mW/m2 to 66.1 mW/m2, which is in the same order of magnitude as previous results. In the calculation, lunar surface heat flow includes the heat flow from the non-uniform distribution of radioactive elements K, U and Th and that from secular cooling of the Moon. The calculation of heat flow from radioactive elements is based on the assumption that the radioactive decay of K, U and Th on the Moon is the same as that on the Earth. The heat flow from secular cooling of the Moon is assumed to be equal to the global average radioactive heat flow. Firstly we construct a relationship between radioactive elements K, U and Th and lunar surface heat flow. The key parameter of the characteristic length scale in the relationship is determined by measured surface heat flow and Th abundances at Apollo 15 and 17 landing sites. Then the distribution of lunar surface heat flow is derived by combining other parameters such as lunar crustal thickness measured by Clementine and lunar crustal density. In addition, correlation analysis of the three radioactive elements is carried out due to the higher resolution of Th abundance and for ease of calculation.

  17. Metal foam heat exchangers for thermal management of fuel cell systems

    NASA Astrophysics Data System (ADS)

    Odabaee, M.; Hooman, K.

    2012-05-01

    The present study explores the possibility of using metal foams for thermal management of fuel cells so that air-cooled fuel cell stacks can be commercialized as replacements for currently-available water-cooled counterparts. Experimental studies have been conducted to examine the heat transfer enhancement from a thin metal foam layer sandwiched between two bipolar plates of a cell. To do this, effects of the key parameters including the free stream velocity and characteristics of metal foam such as porosity, permeability, and form drag coefficient on heat and fluid flow are investigated. The improvements as a result of the application of metal foam layers on fuel cell systems efficiency have been analyzed and discussed. Non-optimized results have shown that to remove the same amount of generated heat, the air-cooled fuel cell systems using aluminum foams require half of the pumping power compared to water-cooled fuel cell systems.

  18. Fuel requirements for low-heat rejection military diesel engines. Interim report, October 1991-September 1993

    SciTech Connect

    Westbrook, S.R.; Stavioha, L.L.; McInnis, L.A.; Likos, W.E.; Yost, D.M.

    1996-01-01

    In the development of high-efficiency advanced engine technology such as low-heat rejection engines and injection systems, the thermal stability of fuel is an important concern. The next generation of engines for combat vehicles will be operating at higher fuel temperatures due to lower waste heat rejection and will be accompanied by higher heat transfer to the fuel injection system. Thus, high-temperature fuel deposit formation is more likely. As a result, two possible methods were evaluated for their potential to reduce fuel deposits: (1) prestress the fuel in an apparatus that feeds the fuel to the engine, or (2) pretreat the fuel with an appropriate additive to reduce deposits in the engine. It was shown that removal of dissolved oxygen from the fuel can significantly reduce the formation of deposits on hot metal surfaces. Prestressing the fuel prior to burning it in the engine was also effective in the reduction of deposit formation. The use of additive pretreatment yielded only limited success.

  19. Use of a commercial heat transfer code to predict horizontally oriented spent fuel rod surface temperatures

    SciTech Connect

    Wix, S.D.; Koski, J.A.

    1993-03-01

    Radioactive spent fuel assemblies are a source of hazardous waste that will have to be dealt with in the near future. It is anticipated that the spent fuel assemblies will be transported to disposal sites in spent fuel transportation casks. In order to design a reliable and safe transportation cask, the maximum cladding temperature of the spent fuel rod arrays must be calculated. A comparison between numerical calculations using commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4{degree}C and 23{degree}C for the low heat dissipation and high heat dissipation, respectively. The temperature predictions using helium as a fill gas are lower for the low and medium heat dissipation levels, but higher at the high heat dissipation. The temperature differences are 1{degree}C and 6{degree}C for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16{degree}C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include experimental uncertainty in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This work demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects will be increasingly important as the amount of dissipated heat increases.

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

    NASA Astrophysics Data System (ADS)

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

    2011-05-01

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

  1. Description of heat flux measurement methods used in hydrocarbon and propellant fuel fires at Sandia.

    SciTech Connect

    Nakos, James Thomas

    2010-12-01

    The purpose of this report is to describe the methods commonly used to measure heat flux in fire applications at Sandia National Laboratories in both hydrocarbon (JP-8 jet fuel, diesel fuel, etc.) and propellant fires. Because these environments are very severe, many commercially available heat flux gauges do not survive the test, so alternative methods had to be developed. Specially built sensors include 'calorimeters' that use a temperature measurement to infer heat flux by use of a model (heat balance on the sensing surface) or by using an inverse heat conduction method. These specialty-built sensors are made rugged so they will survive the environment, so are not optimally designed for ease of use or accuracy. Other methods include radiometers, co-axial thermocouples, directional flame thermometers (DFTs), Sandia 'heat flux gauges', transpiration radiometers, and transverse Seebeck coefficient heat flux gauges. Typical applications are described and pros and cons of each method are listed.

  2. Emission FTIR analyses of thin microscopic patches of jet fuel residues deposited on heated metal surfaces

    NASA Technical Reports Server (NTRS)

    Lauer, J. L.; Vogel, P.

    1986-01-01

    The relationship of fuel stability to fuel composition and the development of mechanisms for deposit formation were investigated. Fuel deposits reduce heat transfer efficiency and increase resistance to fuel flow and are highly detrimental to aircraft performance. Infrared emission Fourier transform spectroscopy was chosen as the primary method of analysis because it was sensitive enough to be used in-situ on tiny patches of monolayers or of only a few molecular layers of deposits which generally proved completely insoluble in any nondestructive solvents. Deposits of four base fuels were compared; dodecane, a dodecane/tetralin blend, commercial Jet A fuel, and a broadened-properties jet fuel particularly rich in polynuclear aromatics. Every fuel in turn was provided with and without small additions of such additives as thiophene, furan, pyrrole, and copper and iron naphthenates.

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

    SciTech Connect

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

    1995-12-31

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

  4. 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.

  5. Mixed-oxide fuel decay heat analysis for BWR LOCA safety evaluation

    SciTech Connect

    Chiang, R. T.

    2013-07-01

    The mixed-oxide (MOX) fuel decay heat behavior is analyzed for Boiling Water Reactor (BWR) Loss of Coolant Accident (LOCA) safety evaluation. The physical reasoning on why the decay heat power fractions of MOX fuel fission product (FP) are significantly lower than the corresponding decay heat power fractions of uranium-oxide (UOX) fuel FP is illustrated. This is primarily due to the following physical phenomena. -The recoverable energies per fission of plutonium (Pu)-239 and Pu-241 are significantly higher than those of uranium (U)-235 and U-238. Consequently, the fission rate required to produce the same amount of power in MOX fuel is significantly lower than that in UOX fuel, which leads to lower subsequent FP generation rate and associated decay heat power in MOX fuel than those in UOX fuel. - The effective FP decay energy per fission of Pu-239 is significantly lower than the corresponding effective FP decay energy per fission of U-235, e.g., Pu-239's 10.63 Mega-electron-Volt (MeV) vs. U-235's 12.81 MeV at the cooling time 0.2 second. This also leads to lower decay heat power in MOX fuel than that in UOX fuel. The FP decay heat is shown to account for more than 90% of the total decay heat immediately after shutdown. The FP decay heat results based on the American National Standard Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979 standard method are shown very close to the corresponding FP decay heat results based on the ANSI/ANS-5.1-2005 standard method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method are shown very close to but mostly slightly lower than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1971 method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method or the ANSI/ANS-5.1-1971 method are shown significantly larger than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1979 standard method or the ANSI/ANS-5.1-2005 standard method. (authors)

  6. Heat transfer analysis of the geologic disposal of spent fuel and high level waste storage canisters

    NASA Astrophysics Data System (ADS)

    Allen, G. K.

    1980-08-01

    Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two and three dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters.

  7. 145. VIEW OF LIQUID NITROGEN/HELIUM HEAT EXCHANGER IN FUEL CONTROL ...

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

    145. VIEW OF LIQUID NITROGEN/HELIUM HEAT EXCHANGER IN FUEL CONTROL ROOM (215), LSB (BLDG. 751), FROM FUEL APRON WITH BAY DOOR OPEN - Vandenberg Air Force Base, Space Launch Complex 3, Launch Pad 3 East, Napa & Alden Roads, Lompoc, Santa Barbara County, CA

  8. Fuel-Flexible Microturbine and Gasifier System for Combined Heat and Power

    SciTech Connect

    2009-12-01

    Capstone Turbine Corporation, in collaboration with the University of California – Irvine, Packer Engineering, and Argonne National Laboratory, will develop and demonstrate a prototype microturbine combined heat and power system fueled by synthesis gas and integrated with a biomass gasifier, enabling reduced fossil fuel consumption and carbon dioxide emissions.

  9. Study on Fuel Cell Network System Considering Reduction in Fuel Cell Capacity Using Load Leveling and Heat Release Loss

    NASA Astrophysics Data System (ADS)

    Obara, Shin'ya; Kudo, Kazuhiko

    Reduction in fuel cell capacity linked to a fuel cell network system is considered. When the power demand of the whole network is small, some of the electric power generated by the fuel cell is supplied to a water electrolysis device, and hydrogen and oxygen gases are generated. Both gases are compressed with each compressor and they are stored in cylinders. When the electric demand of the whole network is large, both gases are supplied to the network, and fuel cells are operated by these hydrogen and oxygen gases. Furthermore, an optimization plan is made to minimize the quantity of heat release of the hot water piping that connects each building. Such an energy network is analyzed assuming connection of individual houses, a hospital, a hotel, a convenience store, an office building, and a factory. Consequently, compared with the conventional system, a reduction of 46% of fuel cell capacity is expected.

  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. Evaluation of a Schatz heat battery on a flexible-fueled vehicle

    NASA Astrophysics Data System (ADS)

    Piotrowski, Gregory K.; Schaefer, Ronald M.

    1991-09-01

    The evaluation is described of a Schatz Heat Battery as a means of reducing cold start emissions from a motor vehicle fueled with both gasoline and M85 high methanol blend fuel. The evaluation was conducted at both 20 and 75 F ambient temperatures. The test vehicle was a flexible fueled 1990 Audi 80 supplied by Volkswagen of America. A description is included of the test vehicle, the test facilities, the analytical methods and test procedures used.

  12. Evaluation of a Schatz heat battery on a flexible-fueled vehicle

    SciTech Connect

    Piotrowski, G.K.; Schaefer, R.M.

    1991-09-01

    The report describes the evaluation of a Schatz Heat Battery as a means of reducing cold start emissions from a motor vehicle fueled with both gasoline and M85 high methanol blend fuel. The evaluation was conducted at both 20 F and 75 F ambient temperatures. The test vehicle was a flexible-fueled 1990 Audi 80 supplied by Volkswagen of America. The report also includes a description of the test vehicle, the test facilities, the analytical methods and test procedures used.

  13. Small oil-fired heating equipment: The effects of fuel quality

    SciTech Connect

    Litzke, W.

    1993-08-01

    The physical and chemical characteristics of fuel can affect its flow, atomization, and combustion, all of which help to define the overall performance of a heating system. The objective of this study was to evaluate the effects of some important parameters of fuel quality on the operation of oil-fired residential heating equipment. The primary focus was on evaluating the effects of the fuel`s sulfur content, aromatics content, and viscosity. Since the characteristics of heating fuel are generally defined in terms of standards (such as ASTM, or state and local fuel-quality requirements), the adequacy and limitations of such specifications also are discussed. Liquid fuels are complex and their properties cannot generally be varied without affecting other properties. To the extent possible, test fuels were specially blended to meet the requirements of the ASTM limits but, at the same time, significant changes were made to the fuels to isolate and vary the selected parameters over broad ranges. A series of combustion tests were conducted using three different types of burners -- a flame-retention head burner, a high static-pressure-retention head burner, and an air-atomized burner. With some adjustments, such modern equipment generally can operate acceptably within a wide range of fuel properties. From the experimental data, the limits of some of the properties could be estimated. The property which most significantly affects the equipment`s performance is viscosity. Highly viscous fuels are poorly atomizated and incompletely burnt, resulting in higher flue gas emissions. Although the sulfur content of the fuel did not significantly affect performance during these short-term studies, other work done at BNL demonstrated that long-term effects due to sulfur can be detrimental in terms of fouling and scale formation on boiler heat exchanger tubes.

  14. A Validation Study of Pin Heat Transfer for MOX Fuel Based on the IFA-597 Experiments

    SciTech Connect

    Phillippe, Aaron M; Clarno, Kevin T; Banfield, James E; Ott, Larry J; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth; Hamilton, Steven P

    2014-01-01

    Abstract The IFA-597 (Integrated Fuel Assessment) experiments from the International Fuel Performance Experiments (IFPE) database were designed to study the thermal behavior of mixed oxide (MOX) fuel and the effects of an annulus on fission gas release in light-water-reactor fuel. An evaluation of nuclear fuel pin heat transfer in the FRAPCON-3.4 and Exnihilo codes for MOX fuel systems was performed, with a focus on the first 20 time steps ( 6 GWd/MT(iHM)) for explicit comparison between the codes. In addition, sensitivity studies were performed to evaluate the effect of the radial power shape and approximations to the geometry to account for the thermocouple hole, dish, and chamfer. The analysis demonstrated relative agreement for both solid (rod 1) and annular (rod 2) fuel in the experiment, demonstrating the accuracy of the codes and their underlying material models for MOX fuel, while also revealing a small energy loss artifact in how gap conductance is currently handled in Exnihilo for chamfered fuel pellets. The within-pellet power shape was shown to significantly impact the predicted centerline temperatures. This has provided an initial benchmarking of the pin heat transfer capability of Exnihilo for MOX fuel with respect to a well-validated nuclear fuel performance code.

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

    SciTech Connect

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

    1983-01-01

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

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

    DOEpatents

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

    2011-03-01

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

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

    SciTech Connect

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

    2013-09-03

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

  18. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    DOEpatents

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

    2010-02-23

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

  19. Theoretical studies of transient criticality of irradiated fuel elements

    SciTech Connect

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

    1987-01-01

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

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

    SciTech Connect

    Chollet, F.

    1984-01-17

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

  1. Single PWR spent fuel assembly heat transfer data for computer code evaluations

    SciTech Connect

    Bates, J.M.

    1986-01-01

    The descriptions and results of two separate heat transfer tests designed to investigate the dry storage of commercial PWR spent fuel assemblies are presented. Presented first are descriptions and selected results from the Fuel Temperature Test performed at the Engine Maintenance and Disassembly facility on the Nevada Test Site. An actual spent fuel assembly from the Turkey Point Unit Number 3 Reactor with a decay heat level of 1.17 KW, was installed vertically in a test stand mounted canister/liner assembly. The boundary temperatures were controlled and the canister backfill gases were alternated between air, helium and vacuum to investigate the primary heat transfer mechanisms of convection, conduction and radiation. The assembly temperature profiles were experimentally measured using installed thermocouple instrumentation. Also presented are the results from the Single Assembly Heat Transfer Test designed and fabricated by Allied General Nuclear Services, under contract to the Department of Energy, and ultimately conducted by the Pacific Northwest Laboratory. For this test, an electrically heated 15 x 15 rod assembly was used to model a single PWR spent fuel assembly. The electrically heated model fuel assembly permitted various ''decay heat'', levels to be tested; 1.0 KW and 0.5 KW were used for these tests. The model fuel assembly was positioned within a prototypic fuel tube and in turn placed within a double-walled sealed cask. The complete test assembly could be positioned at any desired orientation (horizontal, vertical, and 25/sup 0/ from horizontal for the present work) and backfilled as desired (air, helium, or vacuum). Tests were run for all combinations of ''decay heat,'' backfill, and orientation. Boundary conditions were imposed by temperature controlled guard heaters installed on the cask exterior surface.

  2. Performance of Peltier elements as a cryogenic heat flux sensor at temperatures down to 60 K

    NASA Astrophysics Data System (ADS)

    Haruyama, T.

    2001-05-01

    An in situ heat flux measuring technique could be a good tool to investigate the mechanism of radiation heat leak and optimize the performance of multi-layer insulation. There are several types of commercially available heat flux sensors, however, most of these sensors are mainly developed for much higher heat flux measurements, e.g., radiation from an iron furnace, heat leak from LNG tanks to the ground and so on. In cryogenic systems, the typical amount of heat flux from 300 K to the first-stage radiation shield of cryogenic system is around several W/m 2, which is three or four orders of magnitude smaller than that of an iron furnace. A conventional thermoelectric element, known as a Peltier element, has been evaluated as a heat flux sensor at cryogenic temperatures and found to be suitable due to its high output voltage. In this study, the temperature dependence of the sensitivity and thermal resistance of the Peltier elements were investigated at temperatures from 200 down to 60 K for possible practical applications.

  3. Application of fuel cells with heat recovery for integrated utility systems

    NASA Technical Reports Server (NTRS)

    Shields, V.; King, J. M., Jr.

    1975-01-01

    This paper presents the results of a study of fuel cell powerplants with heat recovery for use in an integrated utility system. Such a design provides for a low pollution, noise-free, highly efficient integrated utility. Use of the waste heat from the fuel cell powerplant in an integrated utility system for the village center complex of a new community results in a reduction in resource consumption of 42 percent compared to conventional methods. In addition, the system has the potential of operating on fuels produced from waste materials (pyrolysis and digester gases); this would provide further reduction in energy consumption.

  4. Finite element residual stress analysis of induction heating bended ferritic steel piping

    SciTech Connect

    Kima, Jong Sung; Kim, Kyoung-Soo; Oh, Young-Jin; Chang, Hyung-Young; Park, Heung-Bae

    2014-10-06

    Recently, there is a trend to apply the piping bended by induction heating process to nuclear power plants. Residual stress can be generated due to thermo-mechanical mechanism during the induction heating bending process. It is well-known that the residual stress has important effect on crack initiation and growth. The previous studies have focused on the thickness variation. In part, some studies were performed for residual stress evaluation of the austenitic stainless steel piping bended by induction heating. It is difficult to find the residual stresses of the ferritic steel piping bended by the induction heating. The study assessed the residual stresses of induction heating bended ferriticsteel piping via finite element analysis. As a result, it was identified that high residual stresses are generated on local outersurface region of the induction heating bended ferritic piping.

  5. 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.

  6. Investigating Methods of Heat Recovery from Low-Temperature PEM Fuel Cells in CHP Applications

    SciTech Connect

    Jalalzadeh-Azar, A. A.

    2004-01-01

    Heat recovery from low-temperature proton exchange membrane (PEM) fuel cells poses a number of challenges. In response to these challenges, thermodynamic assessments of proposed heat recovery methods are studied in the context of combined heat and power (CHP) for building applications. Preheating combustion air in conjunction with desiccant dehumidification and absorption cooling technologies is one of the two strategies examined in this study. The other approach integrates the PEM fuel cell with a water-loop heat pump (WLHP) for direct heat recovery. As the primary objective, energy-saving potentials of the adopted heat recovery strategies are estimated with respect to various benchmarks. The quantified energy-saving potentials are translated into effective CHP performance indices and compared with those typically specified by the manufacturers for service hot water applications. The need for developing CHP performance protocols is also discussed in light of the proposed energy recovery techniques - thereby, accomplishing the secondary objective.

  7. Finite Element A Posteriori Error Estimation for Heat Conduction. Degree awarded by George Washington Univ.

    NASA Technical Reports Server (NTRS)

    Lang, Christapher G.; Bey, Kim S. (Technical Monitor)

    2002-01-01

    This research investigates residual-based a posteriori error estimates for finite element approximations of heat conduction in single-layer and multi-layered materials. The finite element approximation, based upon hierarchical modelling combined with p-version finite elements, is described with specific application to a two-dimensional, steady state, heat-conduction problem. Element error indicators are determined by solving an element equation for the error with the element residual as a source, and a global error estimate in the energy norm is computed by collecting the element contributions. Numerical results of the performance of the error estimate are presented by comparisons to the actual error. Two methods are discussed and compared for approximating the element boundary flux. The equilibrated flux method provides more accurate results for estimating the error than the average flux method. The error estimation is applied to multi-layered materials with a modification to the equilibrated flux method to approximate the discontinuous flux along a boundary at the material interfaces. A directional error indicator is developed which distinguishes between the hierarchical modeling error and the finite element error. Numerical results are presented for single-layered materials which show that the directional indicators accurately determine which contribution to the total error dominates.

  8. A bipartite operator interacts with a heat shock element to mediate early meiotic induction of Saccharomyces cerevisiae HSP82

    SciTech Connect

    Szent-Gyorgyi, C.

    1995-12-01

    This report seeks to characterize the activation of meiotic gene in terms of cis-acting DNA elements and their associated factors in Saccharomyces cerevisiae. It was found that vegetative repression and meiotic induction depend on interactions of the promoter-proximal heat shock element with a nearby bipartite repression element. The experiments described explore how two different regulatory pathways induce transcription by stimulating a single classical activation element, a nonspecific heat shock element. 81 refs., 10 figs., 1 tab.

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

    SciTech Connect

    Doane, Harry J.

    1992-07-01

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

  10. Analysis of overall heat balance in self-heated proton-exchange-membrane fuel cells for temperature predictions

    NASA Astrophysics Data System (ADS)

    Koh, Joon-Ho; Hsu, Andrew T.; Akay, Hasan U.; Liou, May-Fun

    The effect of self-heating and cooling by natural convection on a sustainable temperature of PEM fuel cell stacks was studied. Overall mass and heat balance equations are combined to predict self-heated temperatures at various operating conditions. Analyses show that the effect of a heat loss coefficient is more important than other variables such as air flow rate and surrounding temperature. The stack design variables such as active cell area and number of cells also have significant influence on self-controlled temperature. A lower Ohmic resistance of cells is expected to allow a wider range of current load applications. The proposed model can also be used to evaluate heat loss coefficient from measured stack performance and temperature data. Experiments performed on a seven-cell stack of 50 cm 2 active area were used to provide data for the validation of the model.

  11. Fuel and cladding nano-technologies based solutions for long life heat-pipe based reactors

    SciTech Connect

    Popa-Simil, L.

    2012-07-01

    A novel nuclear reactor concept, unifying the fuel pipe with fuel tube functionality has been developed. The structure is a quasi-spherical modular reactor, designed for a very long life. The reactor module unifies the fuel tube with the heat pipe and a graphite beryllium reflector. It also uses a micro-hetero-structure that allows the fission products to be removed in the heat pipe flow and deposited in a getter area in the cold zone of the heat pipe, but outside the neutron flux. The reactor operates as a breed and burn reactor - it contains the fuel pipe with a variable enrichment, starting from the hot-end of the pipe, meant to assure the initial criticality, and reactor start-up followed by area with depleted uranium or thorium that get enriched during the consumption of the first part of the enriched uranium. (authors)

  12. 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

  13. 40 CFR Table C-1 to Subpart C of... - Default CO2 Emission Factors and High Heat Values for Various Types of Fuel

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... Heat Values for Various Types of Fuel C Table C-1 to Subpart C of Part 98 Protection of Environment... Emission Factors and High Heat Values for Various Types of Fuel Default CO2 Emission Factors and High Heat Values for Various Types of Fuel Fuel type Default high heat value Default CO2 emission factor Coal...

  14. 40 CFR Table C-1 to Subpart C of... - Default CO2 Emission Factors and High Heat Values for Various Types of Fuel

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... Heat Values for Various Types of Fuel C Table C-1 to Subpart C of Part 98 Protection of Environment... Emission Factors and High Heat Values for Various Types of Fuel Default CO2 Emission Factors and High Heat Values for Various Types of Fuel Fuel type Default high heat value Default CO2 emission factor Coal...

  15. High-temperature ceramic heat exchanger element for a solar thermal receiver

    NASA Technical Reports Server (NTRS)

    Strumpf, H. J.; Kotchick, D. M.; Coombs, M. G.

    1982-01-01

    A study has been completed on the development of a high-temperature ceramic heat exchanger element to be integrated into a solar reciver producing heated air. A number of conceptual designs were developed for heat exchanger elements of differing configuration. These were evaluated with respect to thermal performance, pressure drop, structural integrity, and fabricability. The final design selection identified a finned ceramic shell as the most favorable concept. The ceramic shell is surrounded by a larger metallic shell. The flanges of the two shells are sealed to provide a leak-tight pressure vessel. The ceramic shell is fabricated by an innovative combination of slip casting the receiver walls and precision casting the heat transfer finned plates. The fins are bonded to the shell during firing. Fabrication of a one-half scale demonstrator ceramic receiver has been completed.

  16. A Review on the Finite Element Methods for Heat Conduction in Functionally Graded Materials

    NASA Astrophysics Data System (ADS)

    Sharma, R.; Jadon, V. K.; Singh, B.

    2015-01-01

    The review presented in this paper focuses mainly on the application of finite element methods for investigating the effect of heat transfer, variation of temperature and other parameters in the functionally graded materials. Different methods have been investigated for thermal conduction in functionally graded materials. The use of FEM for steady state heat transfer has been addressed in this work. The authors have also discussed the utilization of FEM based shear deformation theories and FEM in combination with other methods for the problems involving complexity of the shape and geometry of functionally graded materials. Finite element methods proved to be effective for the solution of heat transfer problem in functionally graded materials. These methods can be used for steady state heat transfer and as well as for transient state.

  17. Northeast Heating Fuel Market The, Assessment and Options

    EIA Publications

    2000-01-01

    In response to the President's request, this study examines how the distillate fuel oil market (and related energy markets) in the Northeast behaved in the winter of 1999-2000, explains the role played by residential, commercial, industrial, and electricity generation sector consumers in distillate fuel oil markets and describes how that role is influenced by the structure of the energy markets in the Northeast

  18. 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.

  19. Improvement of finite element meshes - Heat transfer in an infinite cylinder

    NASA Technical Reports Server (NTRS)

    Kittur, Madan G.; Huston, Ronald L.; Oswald, Fred B.

    1989-01-01

    An extension of a structural finite element mesh improvement technique to heat conduction analysis is presented. The mesh improvement concept was originally presented by Prager in studying tapered, axially loaded bars. It was further shown that an improved mesh can be obtained by minimizing the trace of the stiffnes matrix. These procedures are extended and applied to the analysis of heat conduction in an infinitely long hollow circular cylinder.

  20. Improvement in finite element meshes: Heat transfer in an infinite cylinder

    NASA Technical Reports Server (NTRS)

    Kittur, Madan G.; Huston, Ronald L.; Oswald, Fred B.

    1988-01-01

    An extension of a structural finite element mesh improvement technique to heat conduction analysis is presented. The mesh improvement concept was originally presented by Prager in studying tapered, axially loaded bars. It was further shown that an improved mesh can be obtained by minimizing the trace of the stiffness matrix. These procedures are extended and applied to the analysis of heat conduction in an infinitely long hollow circular cylinder.

  1. Role of fuel upgrading for industry and residential heating

    SciTech Connect

    Merriam, N.W.; Gentile, R.H.

    1995-12-01

    The Koppleman Series C Process is presently being used in pilot plant tests with Wyoming coal to upgrade the Powder River Basin coal containing 30 wt% moisture and having a heating value of 8100 Btu/lb to a product containing less than 1 wt% moisture and having a heating value of 12,200 Btu/lb. This process is described.

  2. Integration Of Heat Transfer Coefficient In Glass Forming Modeling With Special Interface Element

    SciTech Connect

    Moreau, P.; Gregoire, S.; Lochegnies, D.; Cesar de Sa, J.

    2007-05-17

    Numerical modeling of the glass forming processes requires the accurate knowledge of the heat exchange between the glass and the forming tools. A laboratory testing is developed to determine the evolution of the heat transfer coefficient in different glass/mould contact conditions (contact pressure, temperature, lubrication...). In this paper, trials are performed to determine heat transfer coefficient evolutions in experimental conditions close to the industrial blow-and-blow process conditions. In parallel of this work, a special interface element is implemented in a commercial Finite Element code in order to deal with heat transfer between glass and mould for non-meshing meshes and evolutive contact. This special interface element, implemented by using user subroutines, permits to introduce the previous heat transfer coefficient evolutions in the numerical modelings at the glass/mould interface in function of the local temperatures, contact pressures, contact time and kind of lubrication. The blow-and-blow forming simulation of a perfume bottle is finally performed to assess the special interface element performance.

  3. Integration Of Heat Transfer Coefficient In Glass Forming Modeling With Special Interface Element

    NASA Astrophysics Data System (ADS)

    Moreau, P.; César de Sá, J.; Grégoire, S.; Lochegnies, D.

    2007-05-01

    Numerical modeling of the glass forming processes requires the accurate knowledge of the heat exchange between the glass and the forming tools. A laboratory testing is developed to determine the evolution of the heat transfer coefficient in different glass/mould contact conditions (contact pressure, temperature, lubrication…). In this paper, trials are performed to determine heat transfer coefficient evolutions in experimental conditions close to the industrial blow-and-blow process conditions. In parallel of this work, a special interface element is implemented in a commercial Finite Element code in order to deal with heat transfer between glass and mould for non-meshing meshes and evolutive contact. This special interface element, implemented by using user subroutines, permits to introduce the previous heat transfer coefficient evolutions in the numerical modelings at the glass/mould interface in function of the local temperatures, contact pressures, contact time and kind of lubrication. The blow-and-blow forming simulation of a perfume bottle is finally performed to assess the special interface element performance.

  4. Finite-element reentry heat-transfer analysis of space shuttle Orbiter

    NASA Technical Reports Server (NTRS)

    Ko, William L.; Quinn, Robert D.; Gong, Leslie

    1986-01-01

    A structural performance and resizing (SPAR) finite-element thermal analysis computer program was used in the heat-transfer analysis of the space shuttle orbiter subjected to reentry aerodynamic heating. Three wing cross sections and one midfuselage cross section were selected for the thermal analysis. The predicted thermal protection system temperatures were found to agree well with flight-measured temperatures. The calculated aluminum structural temperatures also agreed reasonably well with the flight data from reentry to touchdown. The effects of internal radiation and of internal convection were found to be significant. The SPAR finite-element solutions agreed reasonably well with those obtained from the conventional finite-difference method.

  5. Laminar and turbulent incompressible fluid flow analysis with heat transfer by the finite element method

    SciTech Connect

    Cochran, R.J.

    1992-01-01

    A study of the finite element method applied to two-dimensional incompressible fluid flow analysis with heat transfer is performed using a mixed Galerkin finite element method with the primitive variable form of the model equations. Four biquadratic, quadrilateral elements are compared in this study--the serendipity biquadratic element with bilinear continuous pressure interpolation (Q2(8)-Q1) and the Lagrangian biquadratic element with bilinear continuous pressure interpolation (Q2-Q1) of the Taylor-Hood form. A modified form of the Q2-Q1 element is also studied. The pressure interpolation is augmented by a discontinuous constant shape function for pressure (Q2-Q1+). The discontinuous pressure element formulation makes use of biquadratic shape functions and a discontinuous linear interpolation of the pressure (Q2-P1(3)). Laminar flow solutions, with heat transfer, are compared to analytical and computational benchmarks for flat channel, backward-facing step and buoyancy driven flow in a square cavity. It is shown that the discontinuous pressure elements provide superior solution characteristics over the continuous pressure elements. Highly accurate heat transfer solutions are obtained and the Q2-P1(3) element is chosen for extension to turbulent flow simulations. Turbulent flow solutions are presented for both low turbulence Reynolds number and high Reynolds number formulations of two-equation turbulence models. The following three forms of the length scale transport equation are studied; the turbulence energy dissipation rate ([var epsilon]), the turbulence frequency ([omega]) and the turbulence time scale (tau). It is shown that the low turbulence Reynolds number model consisting of the K - [tau] transport equations, coupled with the damping functions of Shih and Hsu, provides an optimal combination of numerical stability and solution accuracy for the flat channel flow.

  6. Emission FTIR analyses of thin microscopic patches of jet fuel residue deposited on heated metal surface

    NASA Technical Reports Server (NTRS)

    Lauer, J. L.; Vogel, P.

    1984-01-01

    Deposits laid down in patches on metal strips in a high pressure/high temperature fuel system simulator operated with aerated fuel at varying flow rates were analyzed by emission FTIR in terms of functional groups. Significant differences were found in the spectra and amounts of deposits derived from fuels to which small concentrations of oxygen-, nitrogen-, or sulfur-containing heterocyclics or metal naphthenates were added. The spectra of deposits generated on strips by heating fuels and air in a closed container were very different from those of the flowing fluid deposits. One such closed-container dodecane deposit on silver gave a strong surface-enhanced Raman spectrum.

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

    SciTech Connect

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

    1993-04-01

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

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

    SciTech Connect

    Gydesen, S.P.

    1993-07-01

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

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

    NASA Technical Reports Server (NTRS)

    Wallace, Terry C.

    1991-01-01

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

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

    SciTech Connect

    Tolosa, S.C.; Marajofsky, A.

    2004-10-06

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

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

    SciTech Connect

    Olander, D.R.

    1982-01-01

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

  12. An in situ approach to study trace element partitioning in the laser heated diamond anvil cell

    SciTech Connect

    Petitgirard, S.; Mezouar, M.; Borchert, M.; Appel, K.; Liermann, H.-P.; Andrault, D.

    2012-01-15

    Data on partitioning behavior of elements between different phases at in situ conditions are crucial for the understanding of element mobility especially for geochemical studies. Here, we present results of in situ partitioning of trace elements (Zr, Pd, and Ru) between silicate and iron melts, up to 50 GPa and 4200 K, using a modified laser heated diamond anvil cell (DAC). This new experimental set up allows simultaneous collection of x-ray fluorescence (XRF) and x-ray diffraction (XRD) data as a function of time using the high pressure beamline ID27 (ESRF, France). The technique enables the simultaneous detection of sample melting based to the appearance of diffuse scattering in the XRD pattern, characteristic of the structure factor of liquids, and measurements of elemental partitioning of the sample using XRF, before, during and after laser heating in the DAC. We were able to detect elements concentrations as low as a few ppm level (2-5 ppm) on standard solutions. In situ measurements are complimented by mapping of the chemical partitions of the trace elements after laser heating on the quenched samples to constrain the partitioning data. Our first results indicate a strong partitioning of Pd and Ru into the metallic phase, while Zr remains clearly incompatible with iron. This novel approach extends the pressure and temperature range of partitioning experiments derived from quenched samples from the large volume presses and could bring new insight to the early history of Earth.

  13. An in situ approach to study trace element partitioning in the laser heated diamond anvil cell.

    PubMed

    Petitgirard, S; Borchert, M; Andrault, D; Appel, K; Mezouar, M; Liermann, H-P

    2012-01-01

    Data on partitioning behavior of elements between different phases at in situ conditions are crucial for the understanding of element mobility especially for geochemical studies. Here, we present results of in situ partitioning of trace elements (Zr, Pd, and Ru) between silicate and iron melts, up to 50 GPa and 4200 K, using a modified laser heated diamond anvil cell (DAC). This new experimental set up allows simultaneous collection of x-ray fluorescence (XRF) and x-ray diffraction (XRD) data as a function of time using the high pressure beamline ID27 (ESRF, France). The technique enables the simultaneous detection of sample melting based to the appearance of diffuse scattering in the XRD pattern, characteristic of the structure factor of liquids, and measurements of elemental partitioning of the sample using XRF, before, during and after laser heating in the DAC. We were able to detect elements concentrations as low as a few ppm level (2-5 ppm) on standard solutions. In situ measurements are complimented by mapping of the chemical partitions of the trace elements after laser heating on the quenched samples to constrain the partitioning data. Our first results indicate a strong partitioning of Pd and Ru into the metallic phase, while Zr remains clearly incompatible with iron. This novel approach extends the pressure and temperature range of partitioning experiments derived from quenched samples from the large volume presses and could bring new insight to the early history of Earth. PMID:22299967

  14. Air blast and heat radiation from fuel-rich mixture detonations

    NASA Astrophysics Data System (ADS)

    Dorofeev, S. B.; Sidorov, V. P.; Kuznetsov, M. S.; Dvoinishnikov, A. E.; Alekseev, V. I.; Efimenko, A. A.

    1996-06-01

    Large scale experiments were carried out to study the effect fuel concentration on air blast parameters and heart radiation from gaseous detonations. Hemispheric plastic envelope (4 meters in radius) was used with propane-air mixtures containing from 4 to 7 vol. % of fuel. The expressions for overpressures and impulses were determined in Sachs variables. The effect of fuel concentration on blast parameters is shown to be insignificant for the same amount of oxygen in the mixture volume. Thus the blast wave parameters can be described as for stoichiometric mixtures using additional scaling for the explosion energy according to oxygen content (cloud volume). The results of large scale experiments with fuel spray clouds containing 0.16-100 tons of fuel with mean concentration from stoichiometric ( C 0) up to 3 C 0 are reconsidered. These results confirm the proposed scaling of air blast parameters for a wide range of fuel types, cloud volumes and fuel concentrations. Detonations of fuel rich gaseous mixtures result in a strong heat radiation. Heat radiation energy, time and size of the fireball formed are studied as a function of fuel concentration.

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

    SciTech Connect

    Gavalya, R.A.

    1993-09-01

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

  16. 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.

  17. Design and demonstration of a high-temperature, deployable, membrane heat-pipe radiator element

    SciTech Connect

    Trujillo, V.L.; Keddy, E.S.; Merrigan, M.A.

    1989-01-01

    Demonstration of a high-temperature, deployable, membrane heat-pipe radiator element has been conducted. Membrane heat pipes offer the potential for compact storage, ease of transportation, self-deployment, and a high specific radiator performance (kg/kW) for use in thermal reflection systems of space nuclear power plants. A demonstration heat pipe 8-cm wide and 100-cm long was fabricated. The heat pipe containment and wick structure were made of stainless steel and sodium used as the working fluid. The tests demonstrated passive deployment of the high-temperature membrane radiator, simulating a single segment in a flat array, at a temperature of 800 K. Details of test procedures and results of the tests are presented in this paper together with a discussion of the design and development of a full-scale, segmented high-temperature, deployable membrane heat pipe. 5 refs., 7 figs.

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

    SciTech Connect

    De Volpi, A.

    1981-01-01

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

  19. 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.

  20. Charge, mass and heat transfer interactions in solid oxide fuel cells operated with different fuel gases-A sensitivity analysis

    NASA Astrophysics Data System (ADS)

    Nagel, Florian P.; Schildhauer, Tilman J.; Biollaz, Serge M. A.; Stucki, Samuel

    The interaction between charge, heat and mass transfer occurring in SOFCs is investigated applying a finite-volume-based SOFC model. The strong interactions are the consequence of the high degree of integration of different processes (chemical/electrochemical reactions, diffusion, heat and mass transfer) within SOFCs. The understanding of these interactions is a key for the future development and application of SOFCs. The investigation was conducted by means of a sensitivity analysis for two different fuel gases, where one gas features a considerable amount of methane inducing steam reforming reactions as additional disturbance factor in the energy and mass balance system of SOFCs. In order to isolate the impact of the varied model parameters and the according changes in the interactions of charge, mass and heat transfer from side effects, the sensitivity analysis was conducted at constant fuel utilization. It was found that the impact of different fuel gases on the operational conditions of SOFCs dominates geometrical and material-induced phenomena. The power output was most affected by the fuel, followed by the values for the activation polarization activation energy that reflects the employed electrode catalysts activity.

  1. An adaptive finite element method for convective heat transfer with variable fluid properties

    NASA Astrophysics Data System (ADS)

    Pelletier, Dominique; Ilinca, Florin; Hetu, Jean-Francois

    1993-07-01

    This paper presents an adaptive finite element method based on remeshing to solve incompressible viscous flow problems for which fluid properties present a strong temperature dependence. Solutions are obtained in primitive variables using a highly accurate finite element approximation on unstructured grids. Two general purpose error estimators, that take into account fluid properties variations, are presented. The methodology is applied to a problem of practical interest: the thermal convection of corn syrup in an enclosure with localized heating. Predictions are in good agreement with experimental measurements. The method leads to improved accuracy and reliability of finite element predictions.

  2. 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.

  3. Impacts of the Weatherization Assistance Program in fuel-oil heated houses

    SciTech Connect

    Levins, W.P.; Ternes, M.P.

    1994-09-01

    The U.S. DOE Weatherization Assistance Program (WAP) Division requested Oak Ridge National Laboratory to help design and conduct an up-to-date assessment of the Program. The evaluation includes five separate studies; the fuel oil study is the subject of this paper. The primary goal of the fuel-oil study was to provide a region-wide estimate of the space-heating fuel oil saved by the Program in the Northeast during the 1991 and 1992 program years. Other goals include assessing the cost effectiveness of the Program within the fuel-oil submarket, and identifying factors which caused fuel-oil savings to vary. This paper reports only the highlights from the fuel-oil study`s final report.

  4. Three dimensional coupled simulation of thermomechanics, heat, and oxygen diffusion in UO2 nuclear fuel rods

    SciTech Connect

    Chris Newman; Glen Hansen; Derek Gaston

    2009-07-01

    The simulation of nuclear reactor fuel performance involves complex thermomechanical processes between fuel pellets, made of fissile material, and the protective cladding barrier that surrounds the pellets. This paper examines asubset of phenomena that are important in the development of a predictive capability for fuel performance calculations, focusing on thermomechanics and diffusion within UO2 fuel pellets. In this study, correlations from the literature are used for thermal conductivity, specific heat, and oxygen diffusion. This study develops a three dimensional thermomechanical model fully-coupled to an oxygen diffusion model. Both steady state and transient results are examined to compare this three dimensional model with the literature. Further, this equation system is solved in a parallel, fully-coupled, fully-implicit manner using a preconditioned Jacobian-free Newton Krylov method. Numerical results are presented to explore the efficacy of this approach for examining selected fuel performance problems. INL’s BISON fuels performance code is used to perform this analysis.

  5. Modeling the burnout of solid polydisperse fuel under the conditions of external heat transfer

    NASA Astrophysics Data System (ADS)

    Skorik, I. A.; Goldobin, Yu. M.; Tolmachev, E. M.; Gal'perin, L. G.

    2013-11-01

    A self-similar burnout mode of solid polydisperse fuel is considered taking into consideration heat transfer between fuel particles, gases, and combustion chamber walls. A polydisperse composition of fuel is taken into account by introducing particle distribution functions by radiuses obtained for the kinetic and diffusion combustion modes. Equations for calculating the temperatures of particles and gases are presented, which are written for particles average with respect to their distribution functions by radiuses taking into account the fuel burnout ratio. The proposed equations take into consideration the influence of fuel composition, air excess factor, and gas recirculation ratio. Calculated graphs depicting the variation of particle and gas temperatures, and the fuel burnout ratio are presented for an anthracite-fired boiler.

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

    SciTech Connect

    Makenas, B.J.

    1996-03-22

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

  7. Performance of AGR-1 High-Temperature Reactor Fuel During Post-Irradiation Heating Tests

    SciTech Connect

    Morris, Robert Noel; Baldwin, Charles A; Hunn, John D; Demkowicz, Paul; Reber, Edward

    2014-01-01

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide TRISO fuel compacts from the AGR-1 experiment has been evaluated at temperatures of 1600 1800 C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4 to 19.1% FIMA have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 10-6 after 300 h at 1600 C or 100 h at 1800 C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 C, and 85Kr release was very low during the tests (particles with breached SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 C in one compact. Post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

  8. 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

  9. COYOTE: a finite-element computer program for nonlinear heat-conduction problems

    SciTech Connect

    Gartling, D.K.

    1982-10-01

    COYOTE is a finite element computer program designed for the solution of two-dimensional, nonlinear heat conduction problems. The theoretical and mathematical basis used to develop the code is described. Program capabilities and complete user instructions are presented. Several example problems are described in detail to demonstrate the use of the program.

  10. TOPAZ - a finite element heat conduction code for analyzing 2-D solids

    SciTech Connect

    Shapiro, A.B.

    1984-03-01

    TOPAZ is a two-dimensional implicit finite element computer code for heat conduction analysis. This report provides a user's manual for TOPAZ and a description of the numerical algorithms used. Sample problems with analytical solutions are presented. TOPAZ has been implemented on the CRAY and VAX computers.

  11. Finite element procedures for coupled linear analysis of heat transfer, fluid and solid mechanics

    NASA Technical Reports Server (NTRS)

    Sutjahjo, Edhi; Chamis, Christos C.

    1993-01-01

    Coupled finite element formulations for fluid mechanics, heat transfer, and solid mechanics are derived from the conservation laws for energy, mass, and momentum. To model the physics of interactions among the participating disciplines, the linearized equations are coupled by combining domain and boundary coupling procedures. Iterative numerical solution strategy is presented to solve the equations, with the partitioning of temporal discretization implemented.

  12. The Pacific Northwest residential consumer: Perceptions and preferences of home heating fuels, major appliances, and appliance fuels

    SciTech Connect

    Harkreader, S.A.; Hattrup, M.P.

    1988-09-01

    In 1983 the Bonneville Power Administration contracted with the Pacific Northwest Laboratory (PNL) to conduct an analysis of the marketing environment for Bonneville's conservation activities. Since this baseline residential study, PNL has conducted two follow up market research projects: Phase 2 in 1985, and Phase 3, in 1988. In this report the respondents' perceptions, preferences, and fuel switching possibilities of fuels for home heating and major appliances are examined. To aid in effective target marketing, the report identifies market segments according to consumers' demographics, life-cycle, attitudes, and opinions.

  13. Characterising the sintering behaviour of pulverised fuel ash using heating stage microscopy

    SciTech Connect

    Adell, V.; Cheeseman, C.R.; Ferraris, M.; Salvo, M.; Smeacetto, F.; Boccaccini, A.R.

    2007-10-15

    Heating stage microscopy was used to investigate the sintering behaviour of pulverised fuel ash (PFA). The effect of chemical composition, heating rate, maximum temperature and metal inclusions on densification was studied. It was confirmed that dimensional changes of PFA powder compacts can be controlled by selecting appropriate conditions of sintering temperature and heating rate. It was also found that the sintering behaviour of PFA can be modified with the addition of metal inclusions. The results suggest that development of pores and microstructure of lightweight aggregates (LWA) manufactured from PFA can be controlled by changing the key sintering parameters such as temperature, time and heating rate.

  14. Analysis/finite-element combined methodology on temperature distribution of a finite domain with various heat sources

    SciTech Connect

    Wu, H.W.; Shii, Sheng Hwa . Dept. of Naval Architecture and Marine Engineering)

    1994-06-01

    A new method, involving the combined use of analysis and the finite-element method, is applicable to the heat conduction problem with isolated heat sources. Unlike the finite-element method the analysis/finite-element combined method is able to discretize the distributed sources with discontinuities into course elements, and the solution is still calculated accurately. The results are compared in tables with exact solutions and other numerical data, and the agreement is found to be good.

  15. Hybrid transfinite element modeling/analysis of nonlinear heat conduction problems involving phase change

    NASA Technical Reports Server (NTRS)

    Tamma, Kumar K.; Railkar, Sudhir B.

    1988-01-01

    The present paper describes the applicability of hybrid transfinite element modeling/analysis formulations for nonlinear heat conduction problems involving phase change. The methodology is based on application of transform approaches and classical Galerkin schemes with finite element formulations to maintain the modeling versatility and numerical features for computational analysis. In addition, in conjunction with the above, the effects due to latent heat are modeled using enthalpy formulations to enable a physically realistic approximation to be dealt computationally for materials exhibiting phase change within a narrow band of temperatures. Pertinent details of the approach and computational scheme adapted are described in technical detail. Numerical test cases of comparative nature are presented to demonstrate the applicability of the proposed formulations for numerical modeling/analysis of nonlinear heat conduction problems involving phase change.

  16. Finite-element simulation of transient heat response in ultrasonic transducers.

    PubMed

    Ando, E; Kagawa, Y

    1992-01-01

    The application of the finite-element method to a transient heat response problem in electrostrictive ultrasonic transducers during their pulsed operation is described. The temperature and thermal stress distribution are of practical importance for the design of the ultrasonic transducers when they are operated at intense levels. Mechanical vibratory loss is responsible for heat in the elastic parts, while dielectric loss is responsible in the ferroelectric parts. A finite-element computer model is proposed for the temperature change evaluation in the transducers with time. Natural and forced cooling convection and heat radiation from the transducers' boundaries are included. Simulation is made for Langevin-type transducer models, for which comparison is made with experimental data. PMID:18267653

  17. Finite-element simulation of transient heat response in ultrasonic transducers

    NASA Astrophysics Data System (ADS)

    Ando, Ei'ichi; Kagawa, Yukio

    1992-05-01

    The application of the finite-element method to a transient heat response problem in electrostrictive ultrasonic transducers during their pulsed operation is described. The temperature and thermal stress distribution are of practical importance for the design of the ultrasonic transducers when they are operated at intense levels. Mechanical vibratory loss is responsible for heat in the elastic parts while dielectric loss in the ferroelectric parts. A finite-element computer model is proposed for the temperature change evaluation in the transducers with time. Natural and forced cooling convection and heat radiation from the transducers' boundaries are included. Simulation is made for Langevin-type transducer models, for which comparison is made with experimental data.

  18. Gen Purpose 1-D Finite Element Network Fluid Flow Heat Transfer System Simulator

    Energy Science and Technology Software Center (ESTSC)

    1993-08-02

    SAFSIM (System Analysis Flow Simulator) is a FORTRAN computer program to simulate the integrated performance of systems involving fluid mechanics, heat transfer, and reactor dynamics. SAFSIM provides sufficient versatility to allow the engineering simulation of almost any system, from a backyard sprinkler system to a clustered nuclear reactor propulsion system. In addition to versatility, speed and robustness are primary SAFSIM development goals. SAFSIM contains three basic physics modules: (1) a one-dimensional finite element fluid mechanicsmore » module with multiple flow network capability; (2) a one-dimensional finite element structure heat transfer module with multiple convection and radiation exchange capability; and (3) a point reactor dynamics module with reactivity feedback and decay heat capability. SAFSIM can be used for compressible and incompressible, single-phase, multicomponent flow systems.« less

  19. Application of Flame-Sprayed Coatings as Heating Elements for Polymer-Based Composite Structures

    NASA Astrophysics Data System (ADS)

    Lopera-Valle, Adrián; McDonald, André

    2015-10-01

    Flame-sprayed nickel-chromium-aluminum-yttrium (NiCrAlY) and nickel-chromium (NiCr) coatings were deposited on fiber-reinforced polymer composites for use as heating elements of structures that were exposed to cold environments. Electrical current was applied to the coatings to increase the surface temperature by way of Joule heating. The surface temperature profiles of the coatings were measured under free and forced convection conditions at different ambient temperatures, ranging from -25 to 23 °C. It was found that at ambient air temperatures below 0 °C, the surface temperature of the coating remained above 0 °C for both the forced and free convection conditions, and there was a nearly homogeneous temperature distribution over the coating surface. This suggests that flame-sprayed coatings could be used as heating elements to mitigate ice accretion on structures, without the presence of areas of localized high temperature.

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

    SciTech Connect

    Luna, R. E.

    2006-07-01

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

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

    SciTech Connect

    Hai Huang; Ben Spencer; Jason Hales

    2014-10-01

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

  2. Mathematical model of a plate fin heat exchanger operating under solid oxide fuel cell working conditions

    NASA Astrophysics Data System (ADS)

    Kaniowski, Robert; Poniewski, Mieczysław

    2013-12-01

    Heat exchangers of different types find application in power systems based on solid oxide fuel cells (SOFC). Compact plate fin heat exchangers are typically found to perfectly fit systems with power output under 5 kWel. Micro-combined heat and power (micro-CHP) units with solid oxide fuel cells can exhibit high electrical and overall efficiencies, exceeding 85%, respectively. These values can be achieved only when high thermal integration of a system is assured. Selection and sizing of heat exchangers play a crucial role and should be done with caution. Moreover, performance of heat exchangers under variable operating conditions can strongly influence efficiency of the complete system. For that reason, it becomes important to develop high fidelity mathematical models allowing evaluation of heat exchangers under modified operating conditions, in high temperature regimes. Prediction of pressure and temperatures drops at the exit of cold and hot sides are important for system-level studies. Paper presents dedicated mathematical model used for evaluation of a plate fin heat exchanger, operating as a part of micro-CHP unit with solid oxide fuel cells.

  3. Evaluation of resistively heated fuel injection technology to reduce cold start emissions and assist starting/driveaway of a methanol-fueled vehicle

    NASA Astrophysics Data System (ADS)

    Piotrowski, Gregory K.; Schaefer, Ronald M.

    1992-03-01

    The report provides the results from a program to evaluate a set of heated fuel injectors on an M100 fueled vehicle in an attempt to lower cold start emissions of unburned fuel and carbon monoxide and to improve cold start ability and drive ability. This technology was evaluated at several different temperatures.

  4. Partial fuel stratification to control HCCI heat release rates : fuel composition and other factors affecting pre-ignition reactions of two-stage ignition fuels.

    SciTech Connect

    Dec, John E.; Sjoberg, Carl-Magnus G.; Cannella, William; Yang, Yi; Dronniou, Nicolas

    2010-11-01

    Homogeneous charge compression ignition (HCCI) combustion with fully premixed charge is severely limited at high-load operation due to the rapid pressure-rise rates (PRR) which can lead to engine knock and potential engine damage. Recent studies have shown that two-stage ignition fuels possess a significant potential to reduce the combustion heat release rate, thus enabling higher load without knock.

  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. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    SciTech Connect

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  7. A finite element analysis of the freeze/thaw behavior of external artery heat pipes

    NASA Technical Reports Server (NTRS)

    Lu, X. J.; Peterson, G. P.

    1993-01-01

    A two-dimensional finite element model was used to determine the freeze/thaw characteristics of an external artery heat pipe. During startup, the working fluid, which was located in the liquid channel and the circumferential wall grooves, experienced a phase transformation from a solid to a liquid state. The transient heat conduction equations with moving interfacial conditions were solved using the appropriate initial boundary conditions. The modelling results include the cross-sectional temperature distribution and the interfacial or melt front position as a function of time. A fixed grid approach was adopted in the model for the phase-change process during thawing of frozen working fluid. The interfacial position between the liquid and solid regions was found by balancing the latent heat caused by interfacial movement with the heat addition or extraction at the related grid points.

  8. Multiwalled carbon nanotube/polydimethylsiloxane composite films as high performance flexible electric heating elements

    SciTech Connect

    Yan, Jing; Jeong, Young Gyu

    2014-08-04

    High performance elastomeric electric heating elements were prepared by incorporating various contents of pristine multiwalled carbon nanotube (MWCNT) in polydimethylsiloxane (PDMS) matrix by using an efficient solution-casting and curing technique. The pristine MWCNTs were identified to be uniformly dispersed in the PDMS matrix and the electrical percolation of MWCNTs was evaluated to be at ∼0.27 wt. %, where the electrical resistivity of the MWCNT/PDMS composite films dropped remarkably. Accordingly, the composite films with higher MWCNT contents above 0.3 wt. % exhibit excellent electric heating performance in terms of temperature response rapidity and electric energy efficiency at constant applied voltages. In addition, the composite films, which were thermally stable up to 250 °C, showed excellent heating-cooling cyclic performance, which was associated with operational stability in actual electric heating applications.

  9. 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.

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

    NASA Technical Reports Server (NTRS)

    Schneider, Steven J. (Inventor)

    2002-01-01

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

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

    SciTech Connect

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

    2005-05-24

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

  12. Binary Effect of Fly Ash and Palm Oil Fuel Ash on Heat of Hydration Aerated Concrete

    PubMed Central

    Mehmannavaz, Taha; Ismail, Mohammad; Radin Sumadi, Salihuddin; Rafique Bhutta, Muhammad Aamer; Samadi, Mostafa

    2014-01-01

    The binary effect of pulverized fuel ash (PFA) and palm oil fuel ash (POFA) on heat of hydration of aerated concrete was studied. Three aerated concrete mixes were prepared, namely, concrete containing 100% ordinary Portland cement (control sample or Type I), binary concrete made from 50% POFA (Type II), and ternary concrete containing 30% POFA and 20% PFA (Type III). It is found that the temperature increases due to heat of hydration through all the concrete specimens especially in the control sample. However, the total temperature rises caused by the heat of hydration through both of the new binary and ternary concrete were significantly lower than the control sample. The obtained results reveal that the replacement of Portland cement with binary and ternary materials is beneficial, particularly for mass concrete where thermal cracking due to extreme heat rise is of great concern. PMID:24696646

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

    SciTech Connect

    Gohar, Y.

    2000-02-10

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

  14. Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

    NASA Astrophysics Data System (ADS)

    de Wild, P. J.; Nyqvist, R. G.; de Bruijn, F. A.; Stobbe, E. R.

    Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based (e.g. proton exchange membrane fuel cells (PEMFC) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To prevent detrimental effects on the (electro)catalysts in fuel cell-based combined heat and power installations (FC-CHP), sulphur removal from the feedstock is mandatory. An experimental bench-marking study of adsorbents has identified several candidates for the removal of sulphur containing odorants at low temperature. Among these adsorbents a new material has been discovered that offers an economically attractive means to remove TetraHydroThiophene (THT), the main European odorant, from natural gas at ambient temperature. The material is environmentally benign, easy to use and possesses good activity (residual sulphur levels below 20 ppbv) and capacity for the common odorant THT in natural gas. When compared to state-of-the-art metal-promoted active carbon the new material has a THT uptake capacity that is up to 10 times larger, depending on temperature and pressure. Promoted versions of the new material have shown potential for the removal of THT at higher temperatures and/or for the removal of other odorants such as mercaptans from natural gas or from LPG.

  15. FEHMN 1.0: Finite element heat and mass transfer code; Revision 1

    SciTech Connect

    Zyvoloski, G.; Dash, Z.; Kelkar, S.

    1992-05-01

    A computer code is described which can simulate non-isothermal multi-phase multicomponent flow in porous media. It is applicable to natural-state studies of geothermal systems and groundwater flow. The equations of heat and mass transfer for multiphase flow in porous and permeable media are solved sing the finite element method. The permeability and porosity of the medium are allowed to depend on pressure and temperature. The code also has provisions for movable air and water phases and noncoupled tracers; that is, tracer solutions that do not affect the heat and mass transfer solutions. The tracers can be passive or reactive. The code can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. A summary of the equations in the model and the numerical solution procedure are provided in this report. A user`s guide and sample problems are also included. The FEHMN (Finite Element Heat and Mass Nuclear) code, described in this report, is a version of FEHM (Finite Element Heat and Mass, Zyvoloski et al., 1988) developed for the Yucca Mountain Site Characterization Project (YMP). The main use of FEHMN will be to assist in the understanding of flow fields in the saturated zone below the potential Yucca Mountain repository.

  16. Boundary element method applied to a gas-fired pin-fin-enhanced heat pipe

    SciTech Connect

    Andraka, C.E.; Knorovsky, G.A.; Drewien, C.A.

    1998-02-01

    The thermal conduction of a portion of an enhanced surface heat exchanger for a gas fired heat pipe solar receiver was modeled using the boundary element and finite element methods (BEM and FEM) to determine the effect of weld fillet size on performance of a stud welded pin fin. A process that could be utilized by others for designing the surface mesh on an object of interest, performing a conversion from the mesh into the input format utilized by the BEM code, obtaining output on the surface of the object, and displaying visual results was developed. It was determined that the weld fillet on the pin fin significantly enhanced the heat performance, improving the operating margin of the heat exchanger. The performance of the BEM program on the pin fin was measured (as computational time) and used as a performance comparison with the FEM model. Given similar surface element densities, the BEM method took longer to get a solution than the FEM method. The FEM method creates a sparse matrix that scales in storage and computation as the number of nodes (N), whereas the BEM method scales as N{sup 2} in storage and N{sup 3} in computation.

  17. 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.

  18. A constitutive heat shock element-binding factor is immunologically identical to the Ku autoantigen.

    PubMed

    Kim, D; Ouyang, H; Yang, S H; Nussenzweig, A; Burgman, P; Li, G C

    1995-06-23

    Analysis of the heat shock element (HSE)-binding proteins in extracts of rodent cells, during heat shock and their post-heat shock recovery, indicates that the regulation of heat shock response involves a constitutive HSE-binding factor (CHBF), in addition to the heat-inducible heat shock factor HSF1. We purified the CHBF to apparent homogeneity from HeLa cells using column chromatographic techniques including an HSE oligonucleotide affinity column. The purified CHBF consists of two polypeptides with apparent molecular masses of 70 and 86 kDa. Immunoblot and gel mobility shift analysis verify that CHBF is identical or closely related to the Ku autoantigen. The DNA binding characteristics of CHBF to double-stranded or single-stranded DNA are similar to that of Ku autoantigen. In gel mobility shift analysis using purified CHBF and recombinant human HSF1, CHBF competes with HSF1 for the binding of DNA sequences containing HSEs in vitro. Furthermore, when Rat-1 cells were co-transfected with human Ku expression vectors and the hsp70-promoter-driven luciferase reporter gene, thermal induction of luciferase is significantly suppressed relative to cells transfected with only the hsp70-luciferase construct. These data suggest a role of CHBF (or Ku protein) in the regulation of heat response in vivo. PMID:7797514

  19. A new hybrid transfinite element computational methodology for applicability to conduction/convection/radiation heat transfer

    NASA Technical Reports Server (NTRS)

    Tamma, Kumar K.; Railkar, Sudhir B.

    1988-01-01

    This paper describes new and recent advances in the development of a hybrid transfinite element computational methodology for applicability to conduction/convection/radiation heat transfer problems. The transfinite element methodology, while retaining the modeling versatility of contemporary finite element formulations, is based on application of transform techniques in conjunction with classical Galerkin schemes and is a hybrid approach. The purpose of this paper is to provide a viable hybrid computational methodology for applicability to general transient thermal analysis. Highlights and features of the methodology are described and developed via generalized formulations and applications to several test problems. The proposed transfinite element methodology successfully provides a viable computational approach and numerical test problems validate the proposed developments for conduction/convection/radiation thermal analysis.

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

    SciTech Connect

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

    1995-12-31

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

  1. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2012-01-01

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

  3. A comparison of the finite difference and finite element methods for heat transfer calculations

    NASA Technical Reports Server (NTRS)

    Emery, A. F.; Mortazavi, H. R.

    1982-01-01

    The finite difference method and finite element method for heat transfer calculations are compared by describing their bases and their application to some common heat transfer problems. In general it is noted that neither method is clearly superior, and in many instances, the choice is quite arbitrary and depends more upon the codes available and upon the personal preference of the analyst than upon any well defined advantages of one method. Classes of problems for which one method or the other is better suited are defined.

  4. COYOTE II - a finite element computer program for nonlinear heat conduction problems. Part I - theoretical background

    SciTech Connect

    Gartling, D.K.; Hogan, R.E.

    1994-10-01

    The theoretical and numerical background for the finite element computer program, COYOTE II, is presented in detail. COYOTE II is designed for the multi-dimensional analysis of nonlinear heat conduction problems and other types of diffusion problems. A general description of the boundary value problems treated by the program is presented. The finite element formulation and the associated numerical methods used in COYOTE II are also outlined. Instructions for use of the code are documented in SAND94-1179; examples of problems analyzed with the code are provided in SAND94-1180.

  5. Fuel savings with conventional hot water space heating systems by incorporating a natural gas powered heat pump. Preliminary project: Development of heat pump technology

    NASA Astrophysics Data System (ADS)

    Vanheyden, L.; Evertz, E.

    1980-12-01

    Compression type air/water heat pumps were developed for domestic heating systems rated at 20 to 150 kW. The heat pump is driven either by a reciprocating piston or rotary piston engine modified to operate on natural gas. Particular features of natural gas engines as prime movers, such as waste heat recovery and variable speed, are stressed. Two systems suitable for heat pump operation were selected from among five different mass produced car engines and were modified to incorporate reciprocating piston compressor pairs. The refrigerants used are R 12 and R 22. Test rig data transferred to field conditions show that the fuel consumption of conventional boilers can be reduced by 50% and more by the installation of engine driven heat pumps. Pilot heat pumps based on a 1,600 cc reciprocating piston engine were built for heating four two-family houses. Pilot pump operation confirms test rig findings. The service life of rotary piston and reciprocating piston engines was investigated. The tests reveal characteristic curves for reciprocating piston engines and include exhaust composition measurements.

  6. 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.

  7. 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.

  8. 147. EAST END OF LIQUID NITROGEN/HELIUM HEAT EXCHANGER IN FUEL ...

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

    147. EAST END OF LIQUID NITROGEN/HELIUM HEAT EXCHANGER IN FUEL CONTROL ROOM (215), LSB (BLDG. 751), WITH ASSOCIATED PIPING AND VALVES - Vandenberg Air Force Base, Space Launch Complex 3, Launch Pad 3 East, Napa & Alden Roads, Lompoc, Santa Barbara County, CA

  9. Uncertainty Analysis on Heat Transfer Correlations for RP-1 Fuel in Copper Tubing

    NASA Technical Reports Server (NTRS)

    Driscoll, E. A.; Landrum, D. B.

    2004-01-01

    NASA is studying kerosene (RP-1) for application in Next Generation Launch Technology (NGLT). Accurate heat transfer correlations in narrow passages at high temperatures and pressures are needed. Hydrocarbon fuels, such as RP-1, produce carbon deposition (coke) along the inside of tube walls when heated to high temperatures. A series of tests to measure the heat transfer using RP-1 fuel and examine the coking were performed in NASA Glenn Research Center's Heated Tube Facility. The facility models regenerative cooling by flowing room temperature RP-1 through resistively heated copper tubing. A Regression analysis is performed on the data to determine the heat transfer correlation for Nusselt number as a function of Reynolds and Prandtl numbers. Each measurement and calculation is analyzed to identify sources of uncertainty, including RP-1 property variations. Monte Carlo simulation is used to determine how each uncertainty source propagates through the regression and an overall uncertainty in predicted heat transfer coefficient. The implications of these uncertainties on engine design and ways to minimize existing uncertainties are discussed.

  10. Purification and characterization of a heat-shock element binding protein from yeast.

    PubMed Central

    Sorger, P K; Pelham, H R

    1987-01-01

    The promoters of heat shock genes are activated when cells are stressed. Activation is dependent on a specific DNA sequence, the heat-shock element (HSE). We describe the purification to homogeneity of an HSE-binding protein from yeast (Saccharomyces cerevisiae), using sequential chromatography of whole cell extracts on heparin-agarose, calf thymus DNA-Sepharose and an affinity column consisting of a repetitive synthetic HSE sequence coupled to Sepharose. The protein runs as a closely spaced doublet of approximately 150 kd on SDS-polyacrylamide gels; mild proteolysis generates a stable 70-kd fragment which retains DNA binding activity. The relative affinities of the protein for a range of variant HSE sequences correlates with the ability of these sequences to support heat-inducible transcription in vivo, suggesting that this polypeptide is involved in the activation of heat-shock promoters. However, the protein was purified from unshocked yeast, and may therefore represent an unactivated form of heat-shock transcription factor. Study of the purified protein should help to define the mechanistic basis of the heat-shock response. Images Fig. 2. Fig. 4. Fig. 5. Fig. 6. Fig. 7. PMID:3319580

  11. Modelling of automotive fuel droplet heating and evaporation: mathematical tools and approximations

    NASA Astrophysics Data System (ADS)

    Sazhin, Sergei S.; Qubeissi, Mansour Al

    2016-06-01

    New mathematical tools and approximations developed for the analysis of automotive fuel droplet heating and evaporation are summarised. The approach to modelling biodiesel fuel droplets is based on the application of the Discrete Component Model (DCM), while the approach to modelling Diesel fuel droplets is based on the application of the recently developed multi-dimensional quasi-discrete model. In both cases, the models are applied in combination with the Effective Thermal Conductivity/Effective Diffusivity model and the implementation in the numerical code of the analytical solutions to heat transfer and species diffusion equations inside droplets. It is shown that the approximation of biodiesel fuel by a single component leads to under-prediction of droplet evaporation time by up to 13% which can be acceptable as a crude approximation in some applications. The composition of Diesel fuel was simplified and reduced to only 98 components. The approximation of 98 components of Diesel fuel with 15 quasi-components/components leads to under-prediction of droplet evaporation time by about 3% which is acceptable in most engineering applications. At the same time, the approximation of Diesel fuel by a single component and 20 alkane components leads to a decrease in the evaporation time by about 19%, compared with the case of approximation of Diesel fuel with 98 components. The approximation of Diesel fuel with a single alkane quasi-component (C14.763H31.526) leads to under-prediction of the evaporation time by about 35% which is not acceptable even for qualitative analysis of the process. In the case when n-dodecane is chosen as the single alkane component, the above-mentioned under-prediction increases to about 44%.

  12. Low Mode Control of Cryogenic ICF Fuel Layers Using Infrared Heating

    SciTech Connect

    London, R A; Kozioziemski, B J; Marinak, M M; Kerbel, G D; Bittner, D N

    2005-07-06

    Infrared heating has been demonstrated as an effective technique to smooth solid hydrogen layers inside transparent cryogenic inertial confinement fusion capsules. Control of the first two Legendre modes of the fuel thickness perturbations using two infrared beams injected into a hohlraum was predicted by modeling and experimentally demonstrated. In the current work, we use coupled ray tracing and heat transfer simulations to explore a wider range of control of long scale length asymmetries. We demonstrate several scenarios to control the first four Legendre modes in the fuel layer using four beams. With such a system, it appears possible to smooth both short and long scale length fuel thickness variations in transparent indirect drive inertial confinement fusion targets.

  13. A p-version finite element method for steady incompressible fluid flow and convective heat transfer

    NASA Technical Reports Server (NTRS)

    Winterscheidt, Daniel L.

    1993-01-01

    A new p-version finite element formulation for steady, incompressible fluid flow and convective heat transfer problems is presented. The steady-state residual equations are obtained by considering a limiting case of the least-squares formulation for the transient problem. The method circumvents the Babuska-Brezzi condition, permitting the use of equal-order interpolation for velocity and pressure, without requiring the use of arbitrary parameters. Numerical results are presented to demonstrate the accuracy and generality of the method.

  14. Natural element method for radiative heat transfer in a semitransparent medium with irregular geometries

    SciTech Connect

    Zhang, Yong; Yi, Hong-Liang; Tan, He-Ping

    2013-05-15

    This paper develops a numerical solution to the radiative heat transfer problem coupled with conduction in an absorbing, emitting and isotropically scattering medium with the irregular geometries using the natural element method (NEM). The walls of the enclosures, having temperature and mixed boundary conditions, are considered to be opaque, diffuse as well as gray. The NEM as a meshless method is a new numerical scheme in the field of computational mechanics. Different from most of other meshless methods such as element-free Galerkin method or those based on radial basis functions, the shape functions used in NEM are constructed by the natural neighbor interpolations, which are strictly interpolant and the essential boundary conditions can be imposed directly. The natural element solutions in dealing with the coupled heat transfer problem for the mixed boundary conditions have been validated by comparison with those from Monte Carlo method (MCM) generated by the authors. For the validation of the NEM solution to radiative heat transfer in the semicircular medium with an inner circle, the results by NEM have been compared with those reported in the literatures. For pure radiative transfer, the upwind scheme is employed to overcome the oscillatory behavior of the solutions in some conditions. The steady state and transient heat transfer problem combined with radiation and conduction in the semicircular enclosure with an inner circle are studied. Effects of various parameters such as the extinction coefficient, the scattering albedo, the conduction–radiation parameter and the boundary emissivity are analyzed on the radiative and conductive heat fluxes and transient temperature distributions.

  15. Putative cis-Regulatory Elements Associated with Heat Shock Genes Activated During Excystation of Cryptosporidium parvum

    PubMed Central

    Lara, Ana M.; Serrano, Myrna; Sheth, Nihar; Buck, Gregory

    2010-01-01

    Background Cryptosporidiosis is a ubiquitous infectious disease, caused by the protozoan parasites Cryptosporidium hominis and C. parvum, leading to acute, persistent and chronic diarrhea worldwide. Although the complications of this disease can be serious, even fatal, in immunocompromised patients of any age, they have also been found to lead to long term effects, including growth inhibition and impaired cognitive development, in infected immunocompetent children. The Cryptosporidium life cycle alternates between a dormant stage, the oocyst, and a highly replicative phase that includes both asexual vegetative stages as well as sexual stages, implying fine genetic regulatory mechanisms. The parasite is extremely difficult to study because it cannot be cultured in vitro and animal models are equally challenging. The recent publication of the genome sequence of C. hominis and C. parvum has, however, significantly advanced our understanding of the biology and pathogenesis of this parasite. Methodology/Principal Findings Herein, our goal was to identify cis-regulatory elements associated with heat shock response in Cryptosporidium using a combination of in silico and real time RT-PCR strategies. Analysis with Gibbs-Sampling algorithms of upstream non-translated regions of twelve genes annotated as heat shock proteins in the Cryptosporidium genome identified a highly conserved over-represented sequence motif in eleven of them. RT-PCR analyses, described herein and also by others, show that these eleven genes bearing the putative element are induced concurrent with excystation of parasite oocysts via heat shock. Conclusions/Significance Our analyses suggest that occurrences of a motif identified in the upstream regions of the Cryptosporidium heat shock genes represent parts of the transcriptional apparatus and function as stress response elements that activate expression of these genes during excystation, and possibly at other stages in the life cycle of the parasite

  16. Natural element method for radiative heat transfer in a semitransparent medium with irregular geometries

    NASA Astrophysics Data System (ADS)

    Zhang, Yong; Yi, Hong-Liang; Tan, He-Ping

    2013-05-01

    This paper develops a numerical solution to the radiative heat transfer problem coupled with conduction in an absorbing, emitting and isotropically scattering medium with the irregular geometries using the natural element method (NEM). The walls of the enclosures, having temperature and mixed boundary conditions, are considered to be opaque, diffuse as well as gray. The NEM as a meshless method is a new numerical scheme in the field of computational mechanics. Different from most of other meshless methods such as element-free Galerkin method or those based on radial basis functions, the shape functions used in NEM are constructed by the natural neighbor interpolations, which are strictly interpolant and the essential boundary conditions can be imposed directly. The natural element solutions in dealing with the coupled heat transfer problem for the mixed boundary conditions have been validated by comparison with those from Monte Carlo method (MCM) generated by the authors. For the validation of the NEM solution to radiative heat transfer in the semicircular medium with an inner circle, the results by NEM have been compared with those reported in the literatures. For pure radiative transfer, the upwind scheme is employed to overcome the oscillatory behavior of the solutions in some conditions. The steady state and transient heat transfer problem combined with radiation and conduction in the semicircular enclosure with an inner circle are studied. Effects of various parameters such as the extinction coefficient, the scattering albedo, the conduction-radiation parameter and the boundary emissivity are analyzed on the radiative and conductive heat fluxes and transient temperature distributions.

  17. Thermal-hydraulic/heat transfer code development for sphere-pac-fueled LMFBRs. [COBRA-3SP code

    SciTech Connect

    Morris, D.G.

    1980-06-01

    Sphere-pac fuel has received much attention recently in light of the development of proliferation-resistant fuel cycles for the Fast Breeder Reactor Program in the United States. However, for sphere-pac fuel to be a viable alternative to conventional pellet fuel, a means to analyze the thermal behavior of sphere-pac-fueled pin bundles is needed. To meet this need, a thermal-hydraulic/heat transfer computer code has been developed for sphere-pac-fueled fast breeder reactors. The code, COBRA-3SP, is a modified version of COBRA-3M incorporating a three-region sphere-pac fuel pin model which permits fuel restructuring. With COBRA-3SP, steady-state and transient analysis of sphere-pac-fueled pin bundles is possible. The validity of the sphere-pac fuel pin model has been verified using experimental results of irradiated sphere-pac fuel.

  18. A Validation Study of Pin Heat Transfer for UO2 Fuel Based on the IFA-432 Experiments

    SciTech Connect

    Phillippe, Aaron M; Clarno, Kevin T; Banfield, James E; Ott, Larry J; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth; Hamilton, Steven P

    2014-01-01

    The IFA-432 (Integrated Fuel Assessment) experiments from the International Fuel Performance Experiments (IFPE) database were designed to study the effects of gap size, fuel density, and fuel densification on fuel centerline temperature in light-water-reactor fuel. An evaluation of nuclear fuel pin heat transfer in the FRAPCON-3.4 and Exnihilo codes for uranium dioxide (UO$_2$) fuel systems was performed, with a focus on the densification stage (2.2 \\unitfrac{GWd}{MT(UO$_{2}$)}). In addition, sensitivity studies were performed to evaluate the effect of the radial power shape and approximations to the geometry to account for the thermocouple hole. The analysis demonstrated excellent agreement for rods 1, 2, 3, and 5 (varying gap thicknesses and density with traditional fuel), demonstrating the accuracy of the codes and their underlying material models for traditional fuel. For rod 6, which contained unstable fuel that densified an order of magnitude more than traditional, stable fuel, the magnitude of densification was over-predicted and the temperatures were outside of the experimental uncertainty. The radial power shape within the fuel was shown to significantly impact the predicted centerline temperatures, whereas modeling the fuel at the thermocouple location as either annular or solid was relatively negligible. This has provided an initial benchmarking of the pin heat transfer capability of Exnihilo for UO$_2$ fuel with respect to a well-validated nuclear fuel performance code.

  19. Fuel processing in integrated micro-structured heat-exchanger reactors

    NASA Astrophysics Data System (ADS)

    Kolb, G.; Schürer, J.; Tiemann, D.; Wichert, M.; Zapf, R.; Hessel, V.; Löwe, H.

    Micro-structured fuel processors are under development at IMM for different fuels such as methanol, ethanol, propane/butane (LPG), gasoline and diesel. The target application are mobile, portable and small scale stationary auxiliary power units (APU) based upon fuel cell technology. The key feature of the systems is an integrated plate heat-exchanger technology which allows for the thermal integration of several functions in a single device. Steam reforming may be coupled with catalytic combustion in separate flow paths of a heat-exchanger. Reactors and complete fuel processors are tested up to the size range of 5 kW power output of a corresponding fuel cell. On top of reactor and system prototyping and testing, catalyst coatings are under development at IMM for numerous reactions such as steam reforming of LPG, ethanol and methanol, catalytic combustion of LPG and methanol, and for CO clean-up reactions, namely water-gas shift, methanation and the preferential oxidation of carbon monoxide. These catalysts are investigated in specially developed testing reactors. In selected cases 1000 h stability testing is performed on catalyst coatings at weight hourly space velocities, which are sufficiently high to meet the demands of future fuel processing reactors.

  20. Impacts of the Weatherization Assistance Program in fuel-oil heated houses

    SciTech Connect

    Levins, W.P.; Ternes, M.P.

    1994-10-01

    In 1990, the US Department of Energy (DOE) initiated a national evaluation of its lowincome Weatherization Assistance Program. This report, which is one of five parts of that evaluation, evaluates the energy savings and cost-effectiveness of the Program as it had been applied to single-family houses heated primarily by fuel-oil. The study was based upon a representative sample (41 local weatherization agencies, 222 weatherized and 115 control houses) from the nine northeastern states during 1991 and 1992 program years. Dwelling-specific and agency-level data on measures installed, costs, and service delivery procedures were collected from the sampled agencies. Space-heating fuel-oil consumption, indoor temperature, and outdoor temperature were monitored at each house. Dwelling characteristics, air-leakage measurements, space-heating system steady-state efficiency measurements, safety inspections, and occupant questionnaires were also collected or performed at each monitored house. We estimate that the Program weatherized a total of 23,400 single-family fuel-oil heated houses in the nine northeastern states during program years 1991 and 1992. Annual fuel-oil savings were calculated using regression techniques to normalize the savings to standard weather conditions. For the northeast region, annual net fuel-oil savings averaged 160 gallons per house, or 17.7% of pre-weatherization consumption. Although indoor temperatures changed in individual houses following weatherization, there was no average change and no significant difference as compared to the control houses; thus, there was no overall indoor temperature takeback effect influencing fuel-oil savings. The weatherization work was performed cost effectively in these houses from the Program perspective, which included both installation costs and overhead and management costs but did not include non-energy benefits (such as employment and environmental).

  1. Functional conservation of cis-regulatory elements of heat-shock genes over long evolutionary distances.

    PubMed

    He, Zhengying; Eichel, Kelsie; Ruvinsky, Ilya

    2011-01-01

    Transcriptional control of gene regulation is an intricate process that requires precise orchestration of a number of molecular components. Studying its evolution can serve as a useful model for understanding how complex molecular machines evolve. One way to investigate evolution of transcriptional regulation is to test the functions of cis-elements from one species in a distant relative. Previous results suggested that few, if any, tissue-specific promoters from Drosophila are faithfully expressed in C. elegans. Here we show that, in contrast, promoters of fly and human heat-shock genes are upregulated in C. elegans upon exposure to heat. Inducibility under conditions of heat shock may represent a relatively simple "on-off" response, whereas complex expression patterns require integration of multiple signals. Our results suggest that simpler aspects of regulatory logic may be retained over longer periods of evolutionary time, while more complex ones may be diverging more rapidly. PMID:21799932

  2. Epigenetic regulation of repetitive elements is attenuated by prolonged heat stress in Arabidopsis.

    PubMed

    Pecinka, Ales; Dinh, Huy Q; Baubec, Tuncay; Rosa, Marisa; Lettner, Nicole; Mittelsten Scheid, Ortrun

    2010-09-01

    Epigenetic factors determine responses to internal and external stimuli in eukaryotic organisms. Whether and how environmental conditions feed back to the epigenetic landscape is more a matter of suggestion than of substantiation. Plants are suitable organisms with which to address this question due to their sessile lifestyle and diversification of epigenetic regulators. We show that several repetitive elements of Arabidopsis thaliana that are under epigenetic regulation by transcriptional gene silencing at ambient temperatures and upon short term heat exposure become activated by prolonged heat stress. Activation can occur without loss of DNA methylation and with only minor changes to histone modifications but is accompanied by loss of nucleosomes and by heterochromatin decondensation. Whereas decondensation persists, nucleosome loading and transcriptional silencing are restored upon recovery from heat stress but are delayed in mutants with impaired chromatin assembly functions. The results provide evidence that environmental conditions can override epigenetic regulation, at least transiently, which might open a window for more permanent epigenetic changes. PMID:20876829

  3. Investigation of Instabilities and Heat Transfer Phenomena in Supercritical Fuels at High Heat Flux and Temperatures

    NASA Technical Reports Server (NTRS)

    Linne, Diane L.; Meyer, Michael L.; Braun, Donald C.; Keller, Dennis J.

    2000-01-01

    A series of heated tube experiments was performed to investigate fluid instabilities that occur during heating of supercritical fluids. In these tests, JP-7 flowed vertically through small diameter tubes at supercritical pressures. Test section heated length, diameter, mass flow rate, inlet temperature, and heat flux were varied in an effort to determine the range of conditions that trigger the instabilities. Heat flux was varied up to 4 BTU/sq in./s, and test section wall temperatures reached as high as 1950 F. A statistical model was generated to explain the trends and effects of the control variables. The model included no direct linear effect of heat flux on the occurrence of the instabilities. All terms involving inlet temperature were negative, and all terms involving mass flow rate were positive. Multiple tests at conditions that produced instabilities provided inconsistent results. These inconsistencies limit the use of the model as a predictive tool. Physical variables that had been previously postulated to control the onset of the instabilities, such as film temperature, velocity, buoyancy, and wall-to-bulk temperature ratio, were evaluated here. Film temperatures at or near critical occurred during both stable and unstable tests. All tests at the highest velocity were stable, but there was no functional relationship found between the instabilities and velocity, or a combination of velocity and temperature ratio. Finally, all of the unstable tests had significant buoyancy at the inlet of the test section, but many stable tests also had significant buoyancy forces.

  4. High-temperature ceramic heat exchanger element for a solar thermal receiver

    NASA Technical Reports Server (NTRS)

    Strumpf, H. J.; Kotchick, D. M.; Coombs, M. G.

    1982-01-01

    A study was performed by AiResearch Manufacturing Company, a division of The Garrett Corporation, on the development a high-temperature ceramic heat exchanger element to be integrated into a solar receiver producing heated air. A number of conceptual designs were developed for heat exchanger elements of differing configuration. These were evaluated with respect to thermal performance, pressure drop, structural integrity, and fabricability. The final design selection identified a finned ceramic shell as the most favorable concept. The shell is surrounded by a larger metallic shell. The flanges of the two shells are sealed to provide a leak-tight pressure vessel. The ceramic shell is to be fabricated by an innovative combination of slip casting the receiver walls and precision casting the heat transfer finned plates. The fins are bonded to the shell during firing. The unit is sized to produce 2150 F ar at 2.7 atm pressure, with a pressure drop of about 2 percent of the inlet pressure. This size is compatible with a solar collector providing a receiver input of 85 kw(th). Fabrication of a one-half scale demonstrator ceramic receiver has been completed.

  5. A high temperature ceramic heat exchanger element for a solar thermal receiver

    NASA Technical Reports Server (NTRS)

    Strumpf, H. J.; Kotchick, D. M.; Coombs, M. G.

    1982-01-01

    The development of a high-temperature ceramic heat exchanger element to be integrated into a solar receiver producing heated air was studied. A number of conceptual designs were developed for heat exchanger elements of differing configuration. These were evaluated with respect to thermal performance, pressure drop, structural integrity, and fabricability. The final design selection identified a finned ceramic shell as the most favorable concept. The shell is surrounded by a larger metallic shell. The flanges of the two shells are sealed to provide a leak-tight pressure vessel. The ceramic shell is to be fabricated by a innovative combination of slip casting the receiver walls and precision casting the heat transfer finned plates. The fins are bonded to the shell during firing. The unit is sized to produce 2150 F air at 2.7 atm pressure, with a pressure drop of about 2 percent of the inlet pressure. This size is compatible with a solar collector providing a receiver input of 85 kw(th). Fabrication of a one-half scale demonstrator ceramic receiver was completed.

  6. Laminar flow of viscoelastic fluids in rectangular ducts with heat transfer: A finite element analysis

    SciTech Connect

    Syrjaelae, S.

    1998-02-01

    A numerical study on the laminar flow and heat transfer behavior of viscoelastic fluids in rectangular ducts is conducted using the finite element approach. A Criminale-Ericksen-Fibley relation is applied to describe the viscoelastic character of the fluid, and a hydrodynamically and thermally fully developed flow with the H1 thermal boundary condition is considered. The finite element procedure employed yields essentially mesh-independent predictions with a fairly moderate computational effort. Computed results are presented and discussed in terms of the secondary flow field, the temperature field, the friction factor and the Nusselt number. In particular it is shown that the presence of a secondary flow markedly alters the temperature field and results in a substantial heat transfer enhancement with all duct aspect ratios considered. The significant heat transfer enhancement as a consequence of fluid elasticity, with virtually no pressure drop increase, is an interesting phenomenon that certainly has application potential in various industrial processes involving fluid flow and heat transfer.

  7. Deposit formation in hydrocarbon rocket fuels with an evaluation of a propane heat transfer correlation

    NASA Technical Reports Server (NTRS)

    Masters, P. A.; Aukerman, C. A.

    1982-01-01

    A high pressure fuel coking testing apparatus was designed and developed and was used to evaluate thermal decomposition limits and carbon decomposition rates in heated copper tubes for hydrocarbon fuels. A commercial propane (90% grade) and chemically pure (CP) propane were tested. Heat transfer to supercritical propane was evaluated at 136 atm, bulk fluid velocities of 6 to 30 m/s, and tube wall temperatures in the range of 422 to 811 K. A forced convection heat transfer correlation developed in a previous test effort verified a prediction of most of the experimental data within a + or - 30% range, with good agreement for the CP propane data. No significant differences were apparent in the predictions derived from the correlation when the carbon resistance was included with the film resistance. A post-test scanning electron microprobe analysis indicated occurrences of migration and interdiffusion of copper into the carbon deposit.

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

    NASA Astrophysics Data System (ADS)

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

    1992-06-01

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

  9. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-05-01

    As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.

  10. Study on Improving Partial Load by Connecting Geo-thermal Heat Pump System to Fuel Cell Network

    NASA Astrophysics Data System (ADS)

    Obara, Shinya; Kudo, Kazuhiko

    Hydrogen piping, the electric power line, and exhaust heat recovery piping of the distributed fuel cells are connected with network, and operational planning is carried out. Reduction of the efficiency in partial load is improved by operation of the geo-thermal heat pump linked to the fuel cell network. The energy demand pattern of the individual houses in Sapporo was introduced. And the analysis method aiming at minimization of the fuel rate by the genetic algorithm was described. The fuel cell network system of an analysis example assumed connecting the fuel cell co-generation of five houses. When geo-thermal heat pump was introduced into fuel cell network system stated in this paper, fuel consumption was reduced 6% rather than the conventional method

  11. A finite element method based microwave heat transfer modeling of frozen multi-component foods

    NASA Astrophysics Data System (ADS)

    Pitchai, Krishnamoorthy

    Microwave heating is fast and convenient, but is highly non-uniform. Non-uniform heating in microwave cooking affects not only food quality but also food safety. Most food industries develop microwavable food products based on "cook-and-look" approach. This approach is time-consuming, labor intensive and expensive and may not result in optimal food product design that assures food safety and quality. Design of microwavable food can be realized through a simulation model which describes the physical mechanisms of microwave heating in mathematical expressions. The objective of this study was to develop a microwave heat transfer model to predict spatial and temporal profiles of various heterogeneous foods such as multi-component meal (chicken nuggets and mashed potato), multi-component and multi-layered meal (lasagna), and multi-layered food with active packages (pizza) during microwave heating. A microwave heat transfer model was developed by solving electromagnetic and heat transfer equations using finite element method in commercially available COMSOL Multiphysics v4.4 software. The microwave heat transfer model included detailed geometry of the cavity, phase change, and rotation of the food on the turntable. The predicted spatial surface temperature patterns and temporal profiles were validated against the experimental temperature profiles obtained using a thermal imaging camera and fiber-optic sensors. The predicted spatial surface temperature profile of different multi-component foods was in good agreement with the corresponding experimental profiles in terms of hot and cold spot patterns. The root mean square error values of temporal profiles ranged from 5.8 °C to 26.2 °C in chicken nuggets as compared 4.3 °C to 4.7 °C in mashed potatoes. In frozen lasagna, root mean square error values at six locations ranged from 6.6 °C to 20.0 °C for 6 min of heating. A microwave heat transfer model was developed to include susceptor assisted microwave heating of a

  12. Advanced development of the boundary element method for steady-state heat conduction

    NASA Technical Reports Server (NTRS)

    Dargush, G. F.; Banerjee, Prasanta K.

    1989-01-01

    Considerable progress has been made in recent years toward advancing the state-of-the-art in solid mechanics boundary element technology. In the present work, much of this new technology is applied in the development of a general-purpose boundary element method (BEM) for steady-state heat conduction. In particular, the BEM implementation involves the use of higher-order conforming elements, self-adaptive integration and multi-region capability. Two- and three-dimensional, as well as axisymmetric analysis, are incorporated within a unified framework. In addition, techniques are introduced for the calculation of boundary flux, and for the inclusion of thermal resistance across interfaces. As a final extension, an efficient formulation is developed for the analysis of solid three-dimensional bodies with embedded holes. For this last class of problems, the new BEM formulation is particularly attractive, since use of the alternatives (i.e. finite element or finite difference methods) is not practical. A number of detailed examples illustrate the suitability and robustness of the present approach for steady-state heat conduction.

  13. Comparisons of hot water microclimate heating and conventional overhead heating on the development and nutritional status of seedling geraniums as well as fuel consumption of these heating regimes

    SciTech Connect

    Schwartz, M.A.

    1984-01-01

    Two crops of seedling geraniums (Pelargonium x hortorum Bailey) were grown for 16 weeks under the microclimate system using ''Gro-Mat'' ethylene propylene diene monomer (EPDM) tubing and a 10/sup 0/C air -21.1/sup 0/ media temperature environment. Another crop was grown with conventional heating at 16.6/sup 0/ air temperature. Evaluations were made on growth, flowering, nutritional status, and energy consumed. The microclimate system sporadically produced a taller plant during the first eight weeks of the October crop compared to the conventional heated crop; and the December crop was affected only twice during the initial eight weeks. This time the microclimate heating system produced a shorter plant. After week eight there were fewer differences in height. One of the principal differences noted in this study was that the conventionally heated plants flowered seven to ten days earlier than the microclimate plants. By the end of the crop (week 16) few differences in elemental concentrations were detected between the two types of heating systems. The microclimate heating system used about 30% less natural gas than did the conventional system. It appears that microclimate heating systems are a possible alternative to conventional heating systems.

  14. Modeling Cladding-Coolant Heat Transfer of High-Burnup Fuel During RIA

    SciTech Connect

    Wenfeng Liu; Kazimi, Mujid S.

    2006-07-01

    This paper describes a model for the cladding-coolant heat transfer of high burnup fuel during a Reactivity Initiated Accident (RIA) which is implemented in the fuel performance code FRAPTRAN 1.2. The minimum stable film boiling temperature, affected by the subcooling and the clad oxidation, is modeled by a modified Henry correlation. This accounts for the effects of thermal properties of the cladding surface on the transient temperature drop during liquid-solid contact. The transition boiling regime is described as the interpolation of the heat flux between two anchor points on the boiling curve: the Critical Heat Flux (CHF) and minimum stable film boiling. The CHF correlation is based on the Zuber hydrodynamic model multiplied by a subcooling factor. Frederking correlation is chosen to model the film boiling regime. The heat conduction through the oxide layer of the cladding surface of high burnup fuel is calculated by solving heat conduction equations with thermal properties of zirconia taken from MATPRO. This model is validated in the FRAPTRAN code for test cases of both high burnup and fresh test fuel rods including the burnup level (0--56 MW d/kg), peak fuel enthalpy deposit (70--190 cal/g), degree of subcooling (0--80 deg. C), and extent of oxidation (0--25 micron). The modified code demonstrates the capability of differentiating between the departure from nucleate boiling (DNB) and none-DNB cases. The predicted peak cladding temperature (PCT) and duration of DNB achieves generally good agreement with the experimental data. It is found that the cladding surface oxidation of high burnup fuel causes an early rewetting of cladding or suppresses DNB due to two factors: 1) Thick zirconia layer may delay the heat conducted to the surface while keeping the surface heat transfer in the most effective nucleate boiling regime. 2) The transient liquid-solid contact resulting from vapor breaking down would cause a lower interface temperature for an oxidized surface

  15. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    PubMed

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment. PMID:16381764

  16. 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

  17. Uncertainty analysis of steady state incident heat flux measurements in hydrocarbon fuel fires.

    SciTech Connect

    Nakos, James Thomas

    2005-12-01

    The objective of this report is to develop uncertainty estimates for three heat flux measurement techniques used for the measurement of incident heat flux in a combined radiative and convective environment. This is related to the measurement of heat flux to objects placed inside hydrocarbon fuel (diesel, JP-8 jet fuel) fires, which is very difficult to make accurately (e.g., less than 10%). Three methods will be discussed: a Schmidt-Boelter heat flux gage; a calorimeter and inverse heat conduction method; and a thin plate and energy balance method. Steady state uncertainties were estimated for two types of fires (i.e., calm wind and high winds) at three times (early in the fire, late in the fire, and at an intermediate time). Results showed a large uncertainty for all three methods. Typical uncertainties for a Schmidt-Boelter gage ranged from {+-}23% for high wind fires to {+-}39% for low wind fires. For the calorimeter/inverse method the uncertainties were {+-}25% to {+-}40%. The thin plate/energy balance method the uncertainties ranged from {+-}21% to {+-}42%. The 23-39% uncertainties for the Schmidt-Boelter gage are much larger than the quoted uncertainty for a radiative only environment (i.e ., {+-}3%). This large difference is due to the convective contribution and because the gage sensitivities to radiative and convective environments are not equal. All these values are larger than desired, which suggests the need for improvements in heat flux measurements in fires.

  18. Laminar and turbulent incompressible fluid flow analysis with heat transfer by the finite element method

    NASA Astrophysics Data System (ADS)

    Cochran, Robert James

    A study of the finite element method applied to two-dimensional incompressible fluid flow analysis with heat transfer is performed using a mixed Galerkin finite element method with the primitive variable form of the model equations. Four biquadratic, quadrilateral elements are compared in this study--the serendipity biquadratic element with bilinear continuous pressure interpolation (Q2(8)-Q1) and the Lagrangian biquadratic element with bilinear continuous pressure interpolation (Q2-Q1) of the Taylor-Hood form. A modified form of the Q-2Q1 element is also studied. The pressure interpolation is augmented by a discontinuous constant shape function for pressure (Q2-Q1+). The discontinuous pressure element formulation makes use of biquadratic shape functions and a discontinuous linear interpolation of the pressure (Q2-P1(3)). Laminar flow solutions, with heat transfer, are compared to analytical and computational benchmarks for flat channel, backward-facing step and buoyancy driven flow in a square cavity. It is shown that the discontinuous pressure elements provide superior solution characteristics over the continuous pressure elements. Highly accurate heat transfer solutions are obtained and the Q2-P1(3) element is chosen for extension to turbulent flow simulations. Turbulent flow solutions are presented for both low turbulence Reynolds number and high Reynolds number formulations of two equation turbulence models. The following three forms of the length scale transport equation are studied: the turbulence energy dissipation rate (epsilon), the turbulence frequency (omega) and the turbulence time scale (tau). It is shown that the low turbulence Reynolds number model consisting of the k-tau transport equations, coupled with the damping functions of Shih and Hsu, provides an optimal combination of numerical stability and solution accuracy for the flat channel flow. Attempts to extend the formulation beyond the flat channel were not successful due to oscillatory

  19. Construction of a composite thin-element TLD using an optical-heating method.

    PubMed

    Yamamoto, O; Yasuno, Y; Minamide, S; Hasegawa, S; Tsutsui, H; Takenaga, M; Yamashita, T

    1982-09-01

    A composite thermoluminescent dosimeter (TLD), composed of four, thin TL elements with a high-speed reader, has been developed by employing an optical-heating method. Each TL element, which is 15 mg/cm2 thick with a 3 mm dia., is prepared by applying Li2B4O7:Cu or CaSO4:Tm to a plastic substrate 14 mg/cm2 thick. Each element can be heated to 350 degree C within 0.8 sec. by IR radiation from a tungsten lamp. The characteristics of this TLD system include the following: (1) the detection limit of the Li2B4O7:Cu is 3 mR and the limit for CaSO4:Tm is 0.1 mR; (2) the energy-dependence curves are similar to the dose-equivalent curve, showing slight under-responses by 15% near 70 KeV for Li2B4O7:Cu and over-responses by 50% at high energies for CaSO4:Tm; (3) despite quick heating, the residual dose is as low as 0.1% of the last exposure signal; (4) responses are very stable for more than 1,000 cycles of repeated exposure readings; (5) no false signal could be detected, even in the cases of sweat or soil contamination; (6) the thin Li2B4O7;Cu element can be used for skin dose monitoring; and (7) the processing time of the automatic reader for the composite dosimeter is 3 hr/500 dosimeters. This TLD system can be applied to personnel dosimetry, gate monitoring and environmental monitoring. PMID:7174331

  20. 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.