Science.gov

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

    2013-01-01

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

  3. Comparative evaluation of fuel element heat conduction models

    SciTech Connect

    Panicker, M.; Dugan, E.T.; Anghaie, S.

    1986-01-01

    Computer codes that predict thermal-hydraulic performance in light water reactors are found to employ a variety of conduction heat transfer models for the determination of the temperature distribution within fuel elements. The objective of this study was to evaluate, in a consistent manner, the relative merits of these various fuel element conduction heat transfer models by comparing accuracy, speed, and computer storage requirements for calculations performed on selected reference or benchmark problems. Methods of particular interest include: (1) implicit finite difference method (FDM) in COBRA-IIIC; (2) weighted residuals method (WRM) in COBRA-IV; (3) nodal integral method (NIM) in TRAC-PF1; and (4) control volume method (CVM) in RELAP5/MOD1.

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

    NASA Astrophysics Data System (ADS)

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

    2017-01-01

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

  5. Natural convection heat transfer analysis of ATR fuel elements

    SciTech Connect

    Langerman, M.A.

    1992-05-01

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

  6. 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)

  7. 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.

  8. 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)

  9. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  10. 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)

  11. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Shackleford, M.H.

    1958-12-16

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

  12. 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)

  13. 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)

  14. JACKETED FUEL ELEMENT

    DOEpatents

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

    1959-02-01

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

  15. 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.

  16. 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.

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

    SciTech Connect

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

    1980-01-01

    Variable width ribbon heating elements have been fabricated which provide a chopped cosine, variable heat flux profile for fuel rod simulators used in test loops by the Breeder Reactor Program Thermal Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor Core Flow Test Loop. Thermal, mechanical, and electrical design considerations result in the derivation of an analytical expression for the ribbon contours. From this, the ribbons are machined and wound on numerically controlled equipment. Postprocessing and inspection results in a wound, variable width ribbon with the precise dimensional, electrical, and mechanical properties needed for use in fuel pin simulators.

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

    NASA Astrophysics Data System (ADS)

    Wilson, Volney C.

    1997-01-01

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

  19. 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.

  20. 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)

  1. 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.

  2. 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)

  3. 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

  4. 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)

  5. 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.

  6. 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.

  7. Surface chemistry effects in finite element modeling of heat transfer in (micron)-fuel cells

    SciTech Connect

    Havstad, M

    2000-12-07

    Equations for modeling surface chemical kinetics by the interaction of gaseous and surface species are presented. The formulation is embedded in a finite element heat transfer code and an ordinary differential equation package is used to solve the surface system of chemical kinetic equations for each iteration within the heat transfer solver. The method is applied to a flow which includes methane and methanol in a microreactor on a chip. A simpler more conventional method, a plug flow reactor model, is then applied to a similar problem. Initial results for steam reforming of methanol are given.

  8. 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)

  9. Compact Fuel Element Environment Test

    NASA Technical Reports Server (NTRS)

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

    2012-01-01

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

  10. FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Carney, K.G. Jr.

    1959-07-14

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

  11. 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.

  12. 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.

  13. 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)

  14. 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)

  15. 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)

  16. FUEL ELEMENT INTERLOCKING ARRANGEMENT

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1963-01-01

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

  17. 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.

  18. 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.

  19. 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)

  20. 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.

  1. 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.

  2. 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)

  3. 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.

  4. 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.

  5. 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.

  6. Resistance heating elements with specific heating profiles

    NASA Technical Reports Server (NTRS)

    Hirschberg, M. H.

    1976-01-01

    Bundled, interrupted, resistance heating elements provide specific heating profiles. Design allows for easily tailored lengths and locations of "hot sections" and larger surface areas for heat radiation.

  7. 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.

  8. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-16

    A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

  9. RECONDITIONING FUEL ELEMENTS

    DOEpatents

    Brandt, H.L.

    1962-02-20

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

  10. FUEL ELEMENTS FOR NUCLEAR REACTORS

    DOEpatents

    Blainey, A.; Lloyd, H.

    1961-07-11

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

  11. 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

  12. 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)

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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)

  18. Spent graphite fuel element processing

    SciTech Connect

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

    1981-07-01

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

  19. 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.

  20. 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)

  1. FUEL ELEMENT AND METHOD OF PREPARATION

    DOEpatents

    Kingston, W.E.

    1961-04-25

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

  2. 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)

  3. 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)

  4. Dryout of BWR fuel elements

    SciTech Connect

    Reisch, Frigyes

    2006-07-01

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

  5. 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.

  6. Monitoring arrangement for vented nuclear fuel elements

    DOEpatents

    Campana, Robert J.

    1981-01-01

    In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

  7. 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)

  8. 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.

  9. Low cost, lightweight fuel cell elements

    NASA Technical Reports Server (NTRS)

    Kindler, Andrew (Inventor)

    2001-01-01

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

  10. Thermionic fuel element technology status

    NASA Technical Reports Server (NTRS)

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

    1985-01-01

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

  11. Fuel elements of research reactors in China

    SciTech Connect

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

    1988-01-01

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

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

    SciTech Connect

    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.

  13. Fuel elements of thermionic converters

    SciTech Connect

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

    1997-01-01

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

  14. Method of monitoring stored nuclear fuel elements

    SciTech Connect

    Borloo, E.; Buergers, W.; Crutzen, S.; Vinche, C.

    1983-05-24

    To monitor a nuclear fuel element or fuel elements located in a store, e.g. a pond in a swimming pool reactor, the store is illuminated ultrasonically using one or more transducers transmitting ultrasonic signals in one or more predetermined directions to obtain an output which, because it depends on the number and relative location of the fuel elements in the store, and the structure of the store itself is distinctive to the fuel elements or elements stored therein. From this distinctive output is derived an identity unique to the stored fuel element or elements and a reference signal indicative of the whole structure when intact, the reference signal and identity being recorded. Subsequent ultrasonic testing of the store and its contents under identical operating conditions produces a signal which is compared to the recorded reference signal and if different therefrom reveals the occurrence of tampering with the store and/or the fuel element or elements.

  15. 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.

  16. 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.

  17. FUEL ELEMENT FOR A NEUTRONIC REACTOR

    DOEpatents

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

    1963-08-13

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

  18. Automated Fuel Element Closure Welding System

    SciTech Connect

    Wahlquist, D.R.

    1993-03-01

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

  19. Automated Fuel Element Closure Welding System

    SciTech Connect

    Wahlquist, D.R.

    1993-01-01

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

  20. 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.

  1. 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

  2. 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)

  3. 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.

  4. Fuel change possibilities in small heat source

    NASA Astrophysics Data System (ADS)

    Durčanský, Peter; Kapjor, Andrej; Jandačka, Jozef

    2017-09-01

    Rural areas are characterized by a larger number of older family houses with higher fuel consumption for heating. Some areas are not gasified, which means that the fuel base for heating the buildings is very limited. Heating is mainly covered by solid fuels with high emissions and low efficiency. But at the same time, the amount of energy in the form of biowaste can be evaluated and used further. We will explore the possibilities to convert biogas to heat of using a gas burner in a small heat source. However, the heat produced can be used other than for heating or hot water production. The added value for heat generation can be the production of electricity, in the use of heat energy through cogeneration unit with unconventional heat engine. The proposed solution could economically benefit the entire system, because electricity is a noble form of energy and its use is versatile.

  5. MRT fuel element inspection at Dounreay

    SciTech Connect

    Gibson, J.

    1997-08-01

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

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

    NASA Astrophysics Data System (ADS)

    Higgs, Jamie D.

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

  7. 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.

  8. Dual-radial cell thermionic fuel element

    NASA Astrophysics Data System (ADS)

    Terrell, Charles W.

    A dual-radial cell thermionic fuel element (TFE) has been proposed and partially evaluated. The cell has the capacity to produce considerably more power per gram of fuel than does a single-cell TFE, with a total electrical power in a fast reactor system of several hundred kWs, conservatively operated.

  9. 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.

  10. 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.

  11. 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.

  12. Nuclear-fuel-cycle education: Module 4. Fuel element design

    SciTech Connect

    Weisman, J.; Eckart, L.

    1981-12-01

    This module briefly reviews the early development of those fuel designs that lead to the selection of the zircaloy-UO/sub 2/ fuel rod which is used in the present generation of light water reactors (LWR). Fuel element design for the LMFBR and for advanced converter reactors will also be presented. The module will emphasize the design characteristics of the zircaloy-UO/sub 2/ fuel rods used in LWR system. To develop a basic understanding of the LWR system, the module will also describe: the UO/sub 2/ fuel rods and assemblies; the thermal and mechanical design properties characteristic of both normal and transient operations; the physical properties of fuel and cladding; the behavior during reactor irradiation of the fuel and cladding; and a simple fuel rod design code applicable with minimum input preparation. Completion of this module should enable the student to prepare a simple preliminary design of a fuel rod for an LWR with the data available by using the analysis techniques presented in the module. Additionally, the student should be prepared to extend this knowledge to other fuel rod design concepts, e.g., those for the LMFBR and for advanced reactor system fuel rods.

  13. 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.

  14. 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.

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

    NASA Technical Reports Server (NTRS)

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

    2011-01-01

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

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Postirradiation examination of thermionic fuel element specimens

    SciTech Connect

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

    1988-12-01

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

  6. 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.

  7. 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.

  8. 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.

  9. Heated transportable fuel cell cartridges

    DOEpatents

    Lance, Joseph R.; Spurrier, Francis R.

    1985-01-01

    A fuel cell stack protective system is made where a plurality of fuel cells, each containing liquid electrolyte subject to crystallization, is enclosed by a containing vessel, and where at least one electric heater is placed in the containing vessel and is capable of preventing electrolyte crystallization.

  10. 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.

  11. 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.

  12. 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.

  13. Liquid fuel injection elements for rocket engines

    NASA Astrophysics Data System (ADS)

    Cox, George B., Jr.

    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 herein 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.

  14. Liquid fuel injection elements for rocket engines

    NASA Astrophysics Data System (ADS)

    Cox, George B., Jr.

    1993-11-01

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

  15. 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.

  16. METHOD OF MAKING WIRE FUEL ELEMENTS

    DOEpatents

    Zambrow, J.L.

    1960-08-01

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

  17. FUEL ELEMENT FOR A NEUTRONIC REACTOR

    DOEpatents

    Tonks, L.

    1959-09-22

    A fuel element is presented for a reactor comprising a stack of conical bodies of nonfissionable material disposed with the vertex up carrying wafers of fissionable material in grooves near the periphery. These bodies are in a jacket which contains a thermally conducting liquid immersing the bodies. Gaseous fission preducts pass upwardly through central apertures in the bodies while fragments of fissionable material are trapped by vertical projections or walls on the upper surface of the bodies.

  18. CONCENTRIC TUBE FUEL ELEMENT SPRING ALIGNMENT SPACER DEVICE

    DOEpatents

    Weems, S.J.

    1963-09-24

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

  19. 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.

  20. (Fuel efficient rapid response water heating module)

    SciTech Connect

    Ripka, C.D.

    1991-02-19

    This patent describes a fuel efficient rapid response water heating module. It comprises: a water heating tank; a burner cavity disposed within the water heating tank; a burner disposed within the burner cavity; means for supplying a mixture of fuel and air to the burner; preheater jacket means, surrounding the water heating tank; flue means disposed within the water heating tank forming a path for the transfer of combustion gases from the burner cavity to the preheater jacket means; means for supplying cold water to the module; outlet means for providing service hot water from the module; preheater means disposed within the preheater jacket means both to preheat water coming into the module from the cold water supply means and to condense combustion gases within the jacket means; means for transferring preheated water from the preheater jacket means to the water heating tank; means, in flow communication with the preheater jacket means, for draining condensate from the preheater jacket means; and means for causing a flow of the fuel and air mixture to the burner and for causing a flow of combustion gases from the burner cavity, through the flue means and through the preheater jacket means to the atmosphere.

  1. 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.

  2. Emission factors of fine particulate matter, organic and elemental carbon, carbon monoxide, and carbon dioxide for four solid fuels commonly used in residential heating by the U.S. Navajo Nation.

    PubMed

    Champion, Wyatt M; Connors, Lea; Montoya, Lupita D

    2017-09-01

    Most homes in the Navajo Nation use wood as their primary heating fuel, often in combination with locally mined coal. Previous studies observed health effects linked to this solid-fuel use in several Navajo communities. Emission factors (EFs) for common fuels used by the Navajo have not been reported using a relevant stove type. In this study, two softwoods (ponderosa pine and Utah juniper) and two high-volatile bituminous coals (Black Mesa and Fruitland) were tested with an in-use residential conventional wood stove (homestove) using a modified American Society for Testing and Materials/U.S. Environmental Protection Agency (ASTM/EPA) protocol. Filter sampling quantified PM2.5 (particulate matter with an aerodynamic diameter ≤2.5 μm) and organic (OC) and elemental (EC) carbon in the emissions. Real-time monitoring quantified carbon monoxide (CO), carbon dioxide (CO2), and total suspended particles (TSP). EFs for these air pollutants were developed and normalized to both fuel mass and energy consumed. In general, coal had significantly higher mass EFs than wood for all pollutants studied. In particular, coal emitted, on average, 10 times more PM2.5 than wood on a mass basis, and 2.4 times more on an energy basis. The EFs developed here were based on fuel types, stove design, and operating protocols relevant to the Navajo Nation, but they could be useful to other Native Nations with similar practices, such as the nearby Hopi Nation. Indoor wood and coal combustion is an important contributor to public health burdens in the Navajo Nation. Currently, there exist no emission factors representative of Navajo homestoves, fuels, and practices. This study developed emission factors for PM2.5, OC, EC, CO, and CO2 using a representative Navajo homestove. These emission factors may be utilized in regional-, national-, and global-scale health and environmental models. Additionally, the protocols developed and results presented here may inform on-going stove design of the

  3. Microfabricated fuel heating value monitoring device

    DOEpatents

    Robinson, Alex L [Albuquerque, NM; Manginell, Ronald P [Albuquerque, NM; Moorman, Matthew W [Albuquerque, NM

    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.

  4. 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.

  5. 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)

  6. 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''.

  7. 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)

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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.

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

    NASA Technical Reports Server (NTRS)

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

    2013-01-01

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

  1. 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.

  2. 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.

  3. NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

    DOEpatents

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

    1957-11-12

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

  4. Assay method for MTR-type fuel elements

    SciTech Connect

    Sher, R.; Shea, P.

    1981-01-01

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

  5. Finite Element Heat & Mass Transfer Code

    SciTech Connect

    Trease, Lynn

    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; and double porosity and double porosity/double permeability capabilities.

  6. FEHM. Finite Element Heat & Mass Transfer Code

    SciTech Connect

    Zyvoloski, G.A.

    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; and double porosity and double porosity/double permeability capabilities.

  7. Pellet-cladding interaction of LMFBR fuel elements at unsteady state. [ISUNE-5 code

    SciTech Connect

    Ma, B.M.

    1981-10-01

    The fuel pellet-cladding interaction (PCI) of liquid-metal fast breeder reactor (LMFBR) fuel elements or fuel rods at unsteady state is analyzed based on experimental results. The heat generation, fuel restructuring, temperature distribution, gap conductance, irradiation swelling, irradiation creep, fuel burnup, fission gas release, fuel pellet cracking, crack healing, cladding cracking, yield failure and fracture failure of the fuel elements are taken into consideration. To improve the sintered (U,Pu)O/sub 2/ fuel performance and reactor core safety at high temperature and fuel burnup, it is desirable to (a) increase and maintain the ductility of cladding material, (b) provide sufficient gap thickness and plenum space for accommodating fission gas release, (c) keep ramps-power increase rate slow and gentle, and (d) reduce the intensity and frequency of transient PCI in order to avoid intense stress fatigue cracking (SFC) and stress corrosion cracking (SCC) due to fission product compounds CsI, CdI/sub 2/, Cs/sub 2/Te, etc. at the inner cladding surface of the fuel elements during PCI. 10 refs.

  8. 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.

  9. 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/.

  10. 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)

  11. 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.

  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. 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.

  14. Nuclear fuel elements and method of making same

    DOEpatents

    Schweitzer, Donald G.

    1992-01-01

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

  15. Design and 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.

  16. Chemomechanical interactions resulting from fuel-alkali metal reactions inside LMFBR oxide fuel elements

    SciTech Connect

    Adamson, M.G.; Vaidyanathan, S.; Bottcher, J.H.; Hofman, G.L.

    1982-01-01

    Chemomechanical interactions inside metal-clad fuel elements are defined as those fuel-cladding mechanical interactions (FCMI) that are influenced by or result from chemical reactions between constituents of the irradiated fuel system. The purpose of the present paper is to interpret some recent experimental and analytical results in terms of chemomechanical reaction mechanisms, with special emphasis on the modeling of breached LMFBR oxide fuel pin behavior.

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

    NASA Technical Reports Server (NTRS)

    Holland, J. W.; Yang, L.

    1984-01-01

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

  18. 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.

  19. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOEpatents

    Flint, O.

    1961-01-10

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

  20. 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

  1. Heating of solar magnetic elements by downflows

    NASA Astrophysics Data System (ADS)

    Hasan, S. S.; Schuessler, M.

    1985-10-01

    The idea that magnetic elements in the photosphere and lower chromosphere of the sun are heated by downflowing gas is quantitatively examined. The time-dependent hydromagnetic equations are solved numerically in the slender flux tube approximation. Viscous terms are retained, and the radiative exchange of heat between the flux tube and the ambient medium are included. Hydrogen ionization and its thermodynamic consequences are treated self-consistently. Starting from a state of hydrostatic and thermal equilibrium, the temporal response due to the onset of a downflow in the tube is studied. After a transient phase lasting a few minutes, a stationary state results that is substantially hotter than the ambient medium over a fairly large height range. Chapman's facular model can be reproduced remarkably well by adjusting the mass flux entering the tube at the upper boundary. The results are comparatively insensitive to viscosity (nu less than or equal to 10 to the 12th sq cm/s), while radiative heat exchange is significant. Some observational implications are discussed, and it is suggested that the necessary mass flux could be provided by overstable oscillations during their downflow phase.

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

    SciTech Connect

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

    2016-01-22

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

  3. 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.

  4. Failed MTR Fuel Element Detect in a Sipping Tests

    SciTech Connect

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

    2004-10-06

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

  5. 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.

  6. 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)

  7. 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.

  8. 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.

  9. 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.

  10. Space reactor fuel element testing in upgraded TREAT

    NASA Astrophysics Data System (ADS)

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

    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 are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

  11. Space reactor fuel element testing in upgraded TREAT

    NASA Astrophysics Data System (ADS)

    Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

    1993-01-01

    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 ˜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 ˜100 MW/L may be achievable.

  12. 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.

  13. 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.

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

    SciTech Connect

    Rest, J.; Hofman, G.L.

    1997-06-01

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

  15. 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.

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

    SciTech Connect

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

    1988-01-01

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

  17. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    SciTech Connect

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M. )

    1992-11-01

    This report discusses three furnace heating tests which were conducted with irradiated, HT9-clad and U-19wt.%Pu-l0wt.%Zr-alloy fuel, Mk-V-type fuel elements in the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory, Illinois. In general, very significant safety margins for fuel-element cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results will be given, as well as discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction found in high-temperature testing of irradiated metallic fuel elements.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

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

    NASA Astrophysics Data System (ADS)

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

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

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. Method and apparatus for diagnosing breached fuel elements

    DOEpatents

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

    1987-03-02

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

  9. 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}.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  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. 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.

  16. 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)

  17. Use of domestic fuels for large-scale space heating and for district heating

    SciTech Connect

    Seppaelae, R.; Asplund, D.

    1980-01-01

    The aim of the study was to survey the heating systems for large-scale space heating and district heating with domestic fuels or under development in Finland, and to study alternative technico-economic applications in the size class of 0.5 - 5 MW.

  18. 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.

  19. 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

  20. 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)

  1. 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)

  2. 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

  3. 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.

  4. 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.

  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. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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)

  13. Technology status of Thermionic Fuel Elements for space nuclear power

    SciTech Connect

    Holland, J.W.; Yang, L.

    1984-08-01

    Over a half-million thermionic test hours were accumulated in the earlier thermionic fuel element (TFE) development program. When the program was terminated in early 1973, TFEs had operated 12,500 h with projected three-year lifetimes, and individual laboratory converters operated more than five years with stable performance. Primary life-limiting factors were thermionic emitter dimensional increases due to interactions with the fuel and electrical insulator structural damage from fast neutrons. Multiple options for extending TFE lifetimes to seven years or longer are available.

  14. 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.

  15. 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.

  16. 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.

  17. Fine fuel heating by radiant flux

    Treesearch

    David Frankman; Brent W. Webb; Bret W. Butler; Don J. Latham

    2010-01-01

    Experiments were conducted wherein wood shavings and Ponderosa pine needles in quiescent air were subjected to a steady radiation heat flux from a planar ceramic burner. The internal temperature of these particles was measured using fine diameter (0.076mm diameter) type K thermocouples. A narrow angle radiometer was used to determine the emissive power generated by the...

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

    DOEpatents

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley, NJ

    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.

  19. 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.

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

    DOEpatents

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

    1959-08-18

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

  1. 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

  2. METHOD AND APPARATUS FOR EXAMINING FUEL ELEMENTS FOR LEAKAGE

    DOEpatents

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

    1963-12-31

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

  3. Consequences of metallic fuel-cladding liquid phase attack during over-temperature transient on fuel element lifetime

    SciTech Connect

    Lahm, C.E.; Koenig, J.F.; Seidel, B.R.

    1990-01-01

    Metallic fuel elements irradiated in EBR-II at temperatures significantly higher than design, causing liquid phase attack of the cladding, were subsequently irradiated at normal operating temperatures to first breach. The fuel element lifetime was compared to that for elements not subjected to the over-temperature transient and found to be equivalent. 1 ref., 3 figs.

  4. 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.

  5. Method for generating heat from waste fuel

    SciTech Connect

    Lamb, F.H.; Lefcort, M.D.; Rada, P.

    1981-11-17

    A combustion method is disclosed in which heat is generated from particulate laden combustible gas containing mineral matter created from gasifying waste wood, coke or other combustible material. The waste is fed into a pile, under-fire combustion air dries and gasifies the waste, oxidizing the fixed carbon in a first chamber to generate heat at a temperature less than the melting temperature of the non-combustible material so as not to form slag. Air is added in the first chamber in an amount less than stoichiometric with the air introduced in a swirling fashion to move the particulate laterally away from the discharge of the primary chamber, impeding the movement of this particulate also by adding secondary combustion air in a downward swirling direction in the secondary chamber so that very little noncombustible material reaches the second chamber where melting can occur.

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

    SciTech Connect

    Yapuncich, F.L.

    1993-03-01

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

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

    SciTech Connect

    Yapuncich, F.L.

    1993-01-01

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

  8. Heat and fuel coupled operation of a high temperature polymer electrolyte fuel cell with a heat exchanger methanol steam reformer

    NASA Astrophysics Data System (ADS)

    Schuller, G.; Vázquez, F. Vidal; Waiblinger, W.; Auvinen, S.; Ribeirinha, P.

    2017-04-01

    In this work a methanol steam reforming (MSR) reactor has been operated thermally coupled to a high temperature polymer electrolyte fuel cell stack (HT-PEMFC) utilizing its waste heat. The operating temperature of the coupled system was 180 °C which is significantly lower than the conventional operating temperature of the MSR process which is around 250 °C. A newly designed heat exchanger reformer has been developed by VTT (Technical Research Center of Finland LTD) and was equipped with commercially available CuO/ZnO/Al2O3 (BASF RP-60) catalyst. The liquid cooled, 165 cm2, 12-cell stack used for the measurements was supplied by Serenergy A/S. The off-heat from the electrochemical fuel cell reaction was transferred to the reforming reactor using triethylene glycol (TEG) as heat transfer fluid. The system was operated up to 0.4 A cm-2 generating an electrical power output of 427 Wel. A total stack waste heat utilization of 86.4% was achieved. It has been shown that it is possible to transfer sufficient heat from the fuel cell stack to the liquid circuit in order to provide the needed amount for vaporizing and reforming of the methanol-water-mixture. Furthermore a set of recommendations is given for future system design considerations.

  9. 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.

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

    PubMed

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

    2017-03-02

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

  11. 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.

  12. General-purpose heat source: Research and development program. Process evaluation, fuel pellet GF-47

    SciTech Connect

    Reimus, M.A.H.; George, T.G.

    1995-12-01

    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. Because the potential for a launch abort or return from orbit exists for any space mission, the heat source must be designed and constructed to survive credible accident environments. Previous testing conducted in support of the Galileo and Ulysses missions has documented the response of the GPHS heat source to a variety of fragment-impact, aging, atmospheric reentry, and Earth-impact conditions. Although heat sources for previous missions were fabricated by the Westinghouse Savannah River Company (WSRC), GPHS fueled-clads required for the Cassini mission to Saturn will be fabricated by Los Alamos National Laboratory (LANL). This evaluation is part of an ongoing program to determine the similarity of GPHS fueled clads and fuel pellets fabricated at LANL to those fabricated at WSRC. Pellet GF-47, which was fabricated at LANL in late 1994, was submitted for chemical and ceramographic analysis. The results indicated that the pellet had a chemical makeup and microstructure within the range of material fabricated at WSRC in the early 1980s.

  13. Investigation of a Tricarbide Grooved Ring Fuel Element for a Nuclear Thermal Rocket

    NASA Technical Reports Server (NTRS)

    Taylor, Brian D.; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

    2017-01-01

    Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the temperature limitations of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment.

  14. 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.

  15. Gamma-ray spectroscopy on irradiated MTR fuel elements

    NASA Astrophysics Data System (ADS)

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

    2000-08-01

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

  16. 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.

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

    SciTech Connect

    Freels, James D; Jain, Prashant K

    2011-01-01

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

  18. A numerical investigation of the influence of radiation and moisture content on pyrolysis and ignition of a leaf-like fuel element

    Treesearch

    B.L. Yashwanth; B. Shotorban; S. Mahalingam; C.W. Lautenberger; David Weise

    2016-01-01

    The effects of thermal radiation and moisture content on the pyrolysis and gas phase ignition of a solid fuel element containing high moisture content were investigated using the coupled Gpyro3D/FDS models. The solid fuel has dimensions of a typical Arctostaphylos glandulosa leaf which is modeled as thin cellulose subjected to radiative heating on...

  19. 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.

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

    SciTech Connect

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

    2013-01-01

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

  1. Nanofoil Heating Elements for Thermal Batteries

    DTIC Science & Technology

    2008-12-01

    microtherm discs are used to limit the axial heat transfer from the stack. 0 100 200 300 400 500 600 700 800 900 1000 1100 -5 5 15 25 35 45 55 65...Time, t / s T em pe ra tu re , θ / °C 3-2-3, 200 lb, no microtherm 3-2-3, 200 lb, microtherm 2-2-2, 200 lb, microtherm 2-3-2, 200 lb, microtherm 2...4-2, 200 lb, microtherm 2-4-2, 100 lb, microtherm 1-4-1, 200 lb, microtherm 1-2-1, 200 lb, microtherm 3-3-3, 200 lb, microtherm

  2. 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.

  3. 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.

  4. 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

  5. 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.

  6. Method of preparing a high heating value fuel product

    SciTech Connect

    Somerville, R.; Fan, L.T.

    1989-10-24

    This patent describes a method of preparing a high heating value fuel product. The method comprising the steps of: blending a high heating value waste material with a cellulosic material; mixing an organic reagent to the blended mixture of the waste material and the cellulosic material, the organic reagent being a mixture having a 4-15 weight percent of a chemical selected from the group consisting of: triethylene, glycol, diethylene glycol, and glycerin propylene glycol; introducing a pozzolanic agent to the blended mixture for controlling the rate of solidification; and forming the blended mixture into a form suitable for handling. Also described is the same method with the mixture of the organic reagent further comprising: a 20-32 weight percent calcium chloride solution. Another method of preparing a fuel product is also described.

  7. 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.

  8. Heat sources in proton exchange membrane (PEM) fuel cells

    NASA Astrophysics Data System (ADS)

    Ramousse, Julien; Lottin, Olivier; Didierjean, Sophie; Maillet, Denis

    In order to model accurately heat transfer in PEM fuel cell, a particular attention had to be paid to the assessment of heat sources in the cell. Although the total amount of heat released is easily computed from its voltage, local heat sources quantification and localization are not simple. This paper is thus a discussion about heat sources/sinks distribution in a single cell, for which many bold assumptions are encountered in the literature. The heat sources or sinks under consideration are: (1) half-reactions entropy, (2) electrochemical activation, (3) water sorption/desorption at the GDL/membrane interfaces, (4) Joule effect in the membrane and (5) water phase change in the GDL. A detailed thermodynamic study leads to the conclusion that the anodic half-reaction is exothermic (Δ Sr ev a = - 226 J mo l-1 K-1) , instead of being athermic as supposed in most of the thermal studies. As a consequence, the cathodic half-reaction is endothermic (Δ Sr ev c = + 62.8 J mo l-1 K-1) , which results in a heat sink at the cathode side, proportional to the current. In the same way, depending on the water flux through the membrane, sorption can create a large heat sink at one electrode and an equivalent heat source at the other. Water phase change in the GDL - condensation/evaporation - results in heat sources/sinks that should also be taken into account. All these issues are addressed in order to properly set the basis of heat transfer modeling in the cell.

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

    NASA Astrophysics Data System (ADS)

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

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

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

    SciTech Connect

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

    1995-12-01

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

  11. Specifications for high flux isotope reactor fuel elements HFIR-FE-3

    SciTech Connect

    Bowden, G.A.; Knight, R.W.

    1984-08-01

    This specification covers requirements for two types of aluminum-base fuel elements which together will be used as the fuel assembly in the High Flux Isotope Reactor (HFIR). Requirements are included for materials of construction, fabrication, assembly, inspection, and quality control to produce fuel elements in accordance with Company drawings.

  12. 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

  13. Trace element diffusion and incorporation in quartz during heating experiments

    NASA Astrophysics Data System (ADS)

    Rottier, B.; Rezeau, H.; Casanova, V.; Kouzmanov, K.; Moritz, R.; Schlöglova, K.; Wälle, M.; Fontboté, L.

    2017-04-01

    Heating of quartz crystals in order to study melt and high-temperature fluid inclusions is a common practice to constrain major physical and chemical parameters of magmatic and hydrothermal processes. Diffusion and modification of trace element content in quartz and its hosted melt inclusions have been investigated through step-heating experiments of both matrix-free quartz crystals and quartz crystals associated with sulfides and other minerals using a Linkam TS1500 stage. Magmatic and hydrothermal quartz were successively analyzed after each heating step for Cu, Al, and Ti using electron probe micro-analyzer. After the last heating step, quartz crystals and their hosted melt inclusions were analyzed by laser ablation inductively coupled plasma mass spectrometry and compared to unheated samples. Heated samples reveal modification of Cu, Li, Na, and B contents in quartz and modification of Cu, Li, Ag, and K concentrations in melt inclusions. Our results show that different mechanisms of Cu, Li, and Na incorporation occur in magmatic and hydrothermal quartz. Heated magmatic quartz records only small, up to a few ppm, enrichment in Cu and Na, mostly substituting for Li. By contrast, heated hydrothermal quartz shows enrichment up to several hundreds of ppm in Cu, Li, and Na, which substitute for originally present H. This study reveals that the composition of both quartz and its hosted melt inclusions may be significantly modified upon heating experiments, leading to erroneous quantification of elemental concentrations. In addition, each quartz crystal also becomes significantly enriched in Cu in the sub-surface layer during heating. We propose that sub-surface Cu enrichment is a direct indication of Cu diffusion in quartz externally sourced from both the surrounding sulfides as well as the copper pins belonging to the heating device. Our study shows that the chemical compositions of both heated quartz and its hosted inclusions must be interpreted with great caution

  14. Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.

    2017-01-01

    To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.

  15. 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.

  16. 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).

  17. 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.

  18. 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.

  19. 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.

  20. An examination of flame shape related to convection heat transfer in deep-fuel beds

    Treesearch

    Kara M. Yedinak; Jack D. Cohen; Jason M. Forthofer; Mark A. Finney

    2010-01-01

    Fire spread through a fuel bed produces an observable curved combustion interface. This shape has been schematically represented largely without consideration for fire spread processes. The shape and dynamics of the flame profile within the fuel bed likely reflect the mechanisms of heat transfer necessary for the pre-heating and ignition of the fuel during fire spread....

  1. Ignition of an organic water-coal fuel droplet floating in a heated-air flow

    NASA Astrophysics Data System (ADS)

    Valiullin, T. R.; Strizhak, P. A.; Shevyrev, S. A.; Bogomolov, A. R.

    2017-01-01

    Ignition of an organic water-coal fuel (CWSP) droplet floating in a heated-air flow has been studied experimentally. Rank B2 brown-coal particles with a size of 100 μm, used crankcase Total oil, water, and a plasticizer were used as the main CWSP components. A dedicated quartz-glass chamber has been designed with inlet and outlet elements made as truncated cones connected via a cylindrical ring. The cones were used to shape an oxidizer flow with a temperature of 500-830 K and a flow velocity of 0.5-5.0 m/s. A technique that uses a coordinate-positioning gear, a nichrome thread, and a cutter element has been developed for discharging CWSP droplets into the working zone of the chamber. Droplets with an initial size of 0.4 to 2.0 mm were used. Conditions have been determined for a droplet to float in the oxidizer flow long enough for the sustainable droplet burning to be initiated. Typical stages and integral ignition characteristics have been established. The integral parameters (ignition-delay times) of the examined processes have been compared to the results of experiments with CWSP droplets suspended on the junction of a quick-response thermocouple. It has been shown that floating fuel droplets ignite much quicker than the ones that sit still on the thermocouple due to rotation of an CWSP droplet in the oxidizer flow, more uniform heating of the droplet, and lack of heat drainage towards the droplet center. High-speed video recording of the peculiarities of floatation of a burning fuel droplet makes it possible to complement the existing models of water-coal fuel burning. The results can be used for a more substantiated modeling of furnace CWSP burning with the ANSYS, Fluent, and Sigma-Flow software packages.

  2. 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

  3. 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

  4. Fuel Composition and Performance Analysis of Endothermically Heated Fuels for Pulse Detonation Engines

    DTIC Science & Technology

    2009-03-01

    68 Figure 32. Improvement of linearity when using heat addition...Turns, 2000:613). In the second zone, known as the induction zone, thermodynamic properties change little. In the induction zone, ideal gas 12...to 600 K (620°F), and then pumped through a 0.01397 m (0.55 in.) ID stainless steel tube inside a resistive heater. As the fuel flowed through the

  5. Release of fission gas during transient heating of LWR fuel

    SciTech Connect

    Gehl, S.M.

    1982-05-01

    The direct electrical heating technique was used to study fission-gas release and mechanical behavior of irradiated light-water reactor (LWR) fuels during thermal transients. An empirical correlation between fission-gas release and transient temperature history was developed for power-cooling mismatch (PCM) and anticipated transients. Gas release during the refill portion of a design-basis loss of cooling accident was estimated to be less than 1%. Fission-gas release during PCM accidents was found to be controlled by intergranular microcracking and the interlinkage of tunnels on grain edges. For high-gas-release transients, the fractional gas release was shown to be equal to the fractional coverage of grain boundaries by microcracks. Temperature calculations indicated that microcracking causes a significant decrease in the fuel thermal conductivity.

  6. FEHM: finite element heat and mass transfer code

    SciTech Connect

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

    1988-03-01

    The finite element heat and mass (FEHM) transfer code is a computer code developed to simulate geothermal and hot dry rock reservoirs. It is also applicable to natural-state studies of geothermal systems and ground-water flow. It solves the equations of heat and mass transfer for multiphase flow in porous and permeable media using the finite element method. The code also has provisions for a noncoupled tracer; that is, the tracer solutions do not affect the heat and mass transfer solutions. It can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. A summary of the equations in the model, the numerical solution procedure, and model verification and validation are provided in this report. A user's guide and sample problems are included in the appendices. 17 refs., 10 figs., 4 tabs.

  7. 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.

  8. 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.

  9. Benchmark Wall Heat Flux Data for a GO2/GH2 Single Element Injector

    NASA Technical Reports Server (NTRS)

    Pal, Sibtosh; Santoro, Robert

    2004-01-01

    Computational Fluid Dynamics (CFD) is becoming an important component of injector design in the rocket industry. Injector designers who use CFD in the design process need to understand the accuracy level of the particular code being used for certain aspects of the design. This paper presents a recent effort to acquire benchmark quality data to be used for CFD code validation. Detailed chamber wall temperature and heat flux data was acquired for a gaseous oxygen, gaseous hydrogen single element shear coaxial injector in a 1.5 inch diameter copper heat sink chamber at Penn State University. The data was taken using both coaxial and water cooled heat flux gauges. Tests were run using hot gases generated from both fuel and oxidizer preburners. Tests were conducted over a chamber pressure range of 300 to 750 psia. Data analysis and uncertainty information will also be presented.

  10. Benchmark Wall Heat Flux Data for a GO2/GH2 Single Element Injector

    NASA Technical Reports Server (NTRS)

    Pal, Sibtosh; Santoro, Robert

    2004-01-01

    Computational Fluid Dynamics (CFD) is becoming an important component of injector design in the rocket industry. Injector designers who use CFD in the design process need to understand the accuracy level of the particular code being used for certain aspects of the design. This paper presents a recent effort to acquire benchmark quality data to be used for CFD code validation. Detailed chamber wall temperature and heat flux data was acquired for a gaseous oxygen, gaseous hydrogen single element shear coaxial injector in a 1.5 inch diameter copper heat sink chamber at Penn State University. The data was taken using both coaxial and water cooled heat flux gauges. Tests were run using hot gases generated from both fuel and oxidizer preburners. Tests were conducted over a chamber pressure range of 300 to 750 psia. Data analysis and uncertainty information will also be presented.

  11. 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.

  12. 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.

  13. 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.

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

    SciTech Connect

    Montierth, Leland M.

    2016-07-19

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

  15. How to estimate recoverable heat energy in wood or bark fuels

    Treesearch

    P. J. Ince

    1979-01-01

    A reference source is provided for estimating the amount of heat energy that may be recovered using wood or bark fuel in typical furnace and boiler or hot air combustion heat recovery systems. A survey of reported data on higher heating values for various species of wood and bark fuels is provided. A set of formulas of a type commonly used by combustion technologists...

  16. 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.

  17. 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.

  18. Nuclear fuel element, and method of forming same

    SciTech Connect

    Taylor, I.N. Jr.; Magee, P.M.

    1990-11-20

    This patent describes improvement to the method of inhibiting an interaction between a metal alloy fissionable fuel for a nuclear reactor and a stainless steel container for the fuel which results in low temperature melting eutectic reaction products of components from the metal alloy fuel and stainless steel. The improvement comprises providing an expandable body of at least one alloying metal for the metal alloy fissionable fuel selected from the group consisting of zirconium, titanium, niobium and molybdenum at least about 2 mils thick interposed into a space between a metal alloy fuel. The fuel consists of metallic uranium and plutonium and the stainless steel container housing the metal alloy fuel therein.

  19. 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).

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

    SciTech Connect

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

    1998-07-01

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

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

    SciTech Connect

    Lawrence, L.A.

    1998-03-13

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

  2. 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.

  3. Steel construction solid fuel firing boiler for civil heating applications

    SciTech Connect

    Jahier, G.

    1985-05-14

    A novel steel construction solid-fuel firing boiler for civil heating applications comprises a space portion defined by side, bottom, cover, front and rear walls forming therein an interspace for water to be heated, a firestone block burner in the space portion separating a fuel containing upper zone and a lower zone where a flame extends and flue gas is conveyed, the burner defining at a top face thereof a housing with a bottom hole for communicating the upper and lower zones and accommodating therein a refractory material body having a structure for holding overlying embers. The housing bottom and the refractory material body define a channel opening at the hole and communicating with an air distribution chamber in the front wall for conveying secondary combustion air. The boiler further comprises an electric fan with an automatic control circuit and primary and secondary combustion air outlet in communication with the upper zone and the distribution chamber, a smoke box, a loading door on the front wall, an inspection door and a communication port between the upper zone and the smoke box closed by a manually operated gate.

  4. 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.

  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. A solid-state heat pump using electrocaloric ceramic elements

    NASA Astrophysics Data System (ADS)

    Hilt, Matthew G.

    The thermoacoustic cycle is a robust thermodynamic cycle that can be generalized to describe and develop an all-solid-state heat pump using generic caloric elements. Ferroelectric barium strontium titanate (BST) and relaxor lead magnesium niobate - lead titanate (PMN-PT) are two candidate materials for the caloric elements using the electrocaloric effect. I developed a procedure to repeatably produce high quality BST and PMN-PT ceramics so that the electrocaloric and dielectric properties could be accurately measured. The measured electrocaloric properties serve as the baseline numbers for calculating the performance of a proposed all-solid-state cooler based on thermoacoustic principles.

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

    NASA Astrophysics Data System (ADS)

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

    1990-04-01

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

  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. 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.

  10. 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.

  11. 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.

  12. 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.

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

    SciTech Connect

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

    1989-01-01

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

  14. 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.

  15. Finite-element technique applied to heat conduction in solids with temperature dependent thermal conductivity

    NASA Technical Reports Server (NTRS)

    Aguirre-Ramirez, G.; Oden, J. T.

    1969-01-01

    Finite element method applied to heat conduction in solids with temperature dependent thermal conductivity, using nonlinear constitutive equation for heat ABCDEFGHIABCDEFGHIABCDEFGHIABCDEFGHIABCDEFGHIABCDEFGHIABCDEFGHIABCDEFGHIABCDEFGHIABCDEFGHIABCDEFGHIABCDEFGH

  16. Efficient linear and nonlinear heat conduction with a quadrilateral element

    NASA Technical Reports Server (NTRS)

    Liu, W. K.; Belytschko, T.

    1983-01-01

    A method is presented for performing efficient and stable finite element calculations of heat conduction with quadrilaterals using one-point quadrature. The stability in space is obtained by using a stabilization matrix which is orthogonal to all linear fields and its magnitude is determined by a stabilization parameter. It is shown that the accuracy is almost independent of the value of the stabilization parameter over a wide range of values; in fact, the values 3, 2, and 1 for the normalized stabilization parameter lead to the 5-point, 9-point finite difference, and fully integrated finite element operators, respectively, for rectangular meshes and have identical rates of convergence in the L2 norm. Eigenvalues of the element matrices, which are needed for stability limits, are also given. Numerical applications are used to show that the method yields accurate solutions with large increases in efficiency, particularly in nonlinear problems.

  17. 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.

  18. Efficient linear and nonlinear heat conduction with a quadrilateral element

    NASA Technical Reports Server (NTRS)

    Liu, W. K.; Belytschko, T.

    1984-01-01

    A method is presented for performing efficient and stable finite element calculations of heat conduction with quadrilaterals using one-point quadrature. The stability in space is obtained by using a stabilization matrix which is orthogonal to all linear fields and its magnitude is determined by a stabilization parameter. It is shown that the accuracy is almost independent of the value of the stabilization parameter over a wide range of values; in fact, the values 3, 2 and 1 for the normalized stabilization parameter lead to the 5-point finite difference, 9-point finite difference and fully integrated finite element operators, respectively, for rectangular meshes; numerical experiments reported here show that the three have identical rates of convergence in the L2 norm. Eigenvalues of the element matrices, which are needed for stability limits, are also given. Numerical applications are used to show that the method yields accurate solutions with large increases in efficiency, particularly in nonlinear problems.

  19. 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.

  20. 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).

  1. Heat transfer in an infinite layer with fractal distribution of heating elements

    NASA Astrophysics Data System (ADS)

    Titov, V.; Stepanov, R.

    2017-06-01

    The effect of inhomogeneous temperature distribution at boundary on convection in an infinite horizontal plane layer is investigated. Direct numerical simulation of the compressible non-isothermal flow in a cubic cell with rigid horizontal walls is performed under periodic vertical boundary conditions. Laminar and weakly non-linear flow regimes are determined. The heated area on the cell bottom obeys regular or fractal distributions. The intensities of heat flux through the layer are compared for different heterogeneous distributions of heating elements at a specific temperature gradient in the case when the area of heated surface remains constant. Fractal geometry shows the appearance of the multiscale structure of the flow and the enhancement of heat transfer.

  2. 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.

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

    SciTech Connect

    Oliver, Brian M.; Klinger, George S.; Abrefah, John; Marschman, Steven C.; MacFarlan, Paul J.; Ritter, Glenn A.

    1999-08-11

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

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

    SciTech Connect

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

    1998-07-01

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

  5. 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.

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

    SciTech Connect

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

    1999-08-11

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

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

    SciTech Connect

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

    1998-08-01

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

  8. 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.

  9. 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

  10. 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.

  11. The reliability of untempered end plug welds on HT9-clad IFR fuel elements

    SciTech Connect

    Crawford, D C; Porter, D L

    1987-02-01

    Welding generally leaves residual stresses in transformed weld zones, which can initiate cracks from flaws already present in the weld zones. When HT9 cools from welding temperatures, a martensite phase forms in the weld fusion zone and heat-affected zone. Because this martensite phase is hard and brittle, it is particularly susceptible to cracking aggravated by residual stresses. This causes concern over the use of untempered welds on HT9-clad fuel elements. To determine if residual stresses present in end-plug weld zones would affect fuel pin performance, HT9 capsules with prototypic TIG- and CD-welded end plugs (in the tempered and as-welded conditions) were pressurized to failure at room temperature, 550{sup 0}C, and 600{sup 0}C. None of the capsules failed in a weld zone. To determine the effects of reactor operating temperatures on untempered welds, prototypic TIG welds were tempered at reactor bulk sodium temperature and an expected sodium outlet temperature for various lengths of time. Subsequent tensile and burst tests of these specimens proved that any embrittling effects that may have been induced in these welds were of no consequence. Hardness tests on longitudinal sections of welds indicated the amount of tempering a weld will receive inreactor after relatively short lengths of time. The pressure burst tests proved that untemperted welds on HT9-clad fuel elements are as reliable as tempered welds; any residual stresses in untempered weld zones were of no consequence. The tempering test showed that welds used in the as-welded condition will sufficiently temper in 7 days at 550{sup 0}C, but will not, sufficiently temper in 7 days at bulk sodium temperature. A comparison of the structure of laser welds to those of CD and TIG welds indicated that untempered laser welds will perform and temper in a manner similar to the TIG welds tested in this effort.

  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. 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.

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

    SciTech Connect

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

    1995-01-01

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

  15. 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.

  16. 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.

  17. 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.

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. Axisymmetric heat conduction analysis under steady state by improved multiple-reciprocity boundary element method

    SciTech Connect

    Ochiai, Yoshihiro

    1995-09-01

    Heat-conduction analysis under steady state without heat generation can easily be treated by the boundary element method. However, in the case with heat conduction with heat generation can approximately be solved without a domain integral by an improved multiple-reciprocity boundary element method. The convention multiple-reciprocity boundary element method is not suitable for complicated heat generation. In the improved multiple-reciprocity boundary element method, on the other hand, the domain integral in each step is divided into point, line, and area integrals. In order to solve the problem, the contour lines of heat generation, which approximate the actual heat generation, are used.

  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. Finite Element Modelling of the Apollo Heat Flow Experiments

    NASA Astrophysics Data System (ADS)

    Platt, J.; Siegler, M. A.; Williams, J.

    2013-12-01

    The heat flow experiments sent on Apollo missions 15 and 17 were designed to measure the temperature gradient of the lunar regolith in order to determine the heat flux of the moon. Major problems in these experiments arose from the fact that the astronauts were not able to insert the probes below the thermal skin depth. Compounding the problem, anomalies in the data have prevented scientists from conclusively determining the temperature dependent conductivity of the soil, which enters as a linear function into the heat flow calculation, thus stymieing them in their primary goal of constraining the global heat production of the Moon. Different methods of determining the thermal conductivity have yielded vastly different results resulting in downward corrections of up to 50% in some cases from the original calculations. Along with problems determining the conductivity, the data was inconsistent with theoretical predictions of the temperature variation over time, leading some to suspect that the Apollo experiment itself changed the thermal properties of the localised area surrounding the probe. The average temperature of the regolith, according to the data, increased over time, a phenomenon that makes calculating the thermal conductivity of the soil and heat flux impossible without knowing the source of error and accounting for it. The changes, possibly resulting from as varied sources as the imprint of the Astronauts boots on the lunar surface, compacted soil around the bore stem of the probe or even heat radiating down the inside of the tube, have convinced many people that the recorded data is unusable. In order to shed some light on the possible causes of this temperature rise, we implemented a finite element model of the probe using the program COMSOL Multi-physics as well as Matlab. Once the cause of the temperature rise is known then steps can be taken to account for the failings of the experiment and increase the data's utility.

  5. Overview of high-temperature fuel behaviour with relevance to CANDU fuel

    NASA Astrophysics Data System (ADS)

    Lewis, B. J.; Iglesias, F. C.; Dickson, R. S.; Williams, A.

    2009-10-01

    This paper provides an overview of high-temperature phenomena in nuclear fuel elements and bundles, with particular relevance to the CANDU fuel design. The paper describes heat generation, fuel thermal response, and thermophysical properties of the fuel and sheath that can affect the thermal and mechanical response of the fuel element. Sources of chemical heat that can arise during accident conditions in the fuel element are also detailed. Specific phenomena associated with fuel restructuring, fuel sheath deformation, fuel-to-sheath heat transfer, fuel sheath failure criteria, oxidation, hydriding and embrittlement of the Zircaloy sheath, gap transport processes in failed elements, fuel/sheath interaction and fuel dissolution by molten cladding are detailed as important phenomena that can impact reactor safety analysis. Fuel behaviour during a power pulse and fuel bundle behaviour that occurs during a severe reactor accident are further considered. The review also points out areas of further research that are needed for a more complete understanding.

  6. A new finite element model for welding heat sources

    NASA Astrophysics Data System (ADS)

    Goldak, John; Chakravarti, Aditya; Bibby, Malcolm

    1984-06-01

    A mathematical model for weld heat sources based on a Gaussian distribution of power density in space is presented. In particular a double ellipsoidal geometry is proposed so that the size and shape of the heat source can be easily changed to model both the shallow penetration arc welding processes and the deeper penetration laser and electron beam processes. In addition, it has the versatility and flexibility to handle non-axisymmetric cases such as strip electrodes or dissimilar metal joining. Previous models assumed circular or spherical symmetry. The computations are performed with ASGARD, a nonlinear transient finite element (FEM) heat flow program developed for the thermal stress analysis of welds.* Computed temperature distributions for submerged arc welds in thick workpieces are compared to the measured values reported by Christensen1 and the FEM calculated values (surface heat source model) of Krutz and Segerlind.2 In addition the computed thermal history of deep penetration electron beam welds are compared to measured values reported by Chong.3 The agreement between the computed and measured values is shown to be excellent.

  7. 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.

  8. 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.

  9. Highly efficient heat recovery system for phosphoric acid fuel cells used for cooling telecommunication equipment

    NASA Astrophysics Data System (ADS)

    Ishizawa, Maki; Okada, Shigeru; Yamashita, Takashi

    To protect the global environment by using energy more efficiently, NTT is developing a phosphoric acid fuel cell (PAFC) energy system for telecommunication cogeneration systems. Fuel cells are used to provide electrical power to telecommunication equipment and the heat energy is used by absorption refrigerators to cool the telecommunication rooms throughout the year. We have recently developed a highly efficient system for recovering heat and water from the exhaust gases of a 200-kW (rated power) fuel cell. It is composed of a shell-and-tube type heat exchanger to recover high-temperature heat and a direct-contact cooler to recover the water efficiently and simply. The reformer and cathode exhaust gases from the fuel cell are first supplied to the heat exchanger and then to the cooler. The high-temperature (85-60°C) heat can be recovered, and the total efficiency including the heat recovered from the fuel-cell stack coolant can be improved by supplying the recovered heat to the dual-heat-input absorption refrigerator. The water needed for operating the fuel cell is also recovered from the exhaust gases. We are currently applying this heat and water recovery system to the PC25C-type fuel cell. Maximum total efficiency including electrical power efficiency is estimated to be 78% at the rated power of 200 kW: composed of 17% heat recovery for the fuel-cell stack coolant, 21% from the exhaust gas by improving the heat exchanger, and 40% from electrical conversion. Next, we plan to evaluate the usefulness of this heat recovery system for cooling telecommunication equipment.

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

    NASA Astrophysics Data System (ADS)

    Mougin, L. J.

    1983-07-01

    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.

  13. Thermal-Hydraulic Bases for the Safety Limits and Limiting Safety System Settings for HFIR Operation at 100 MW and 468 psig Primary Pressure, Using Specially Selected Fuel Elements

    SciTech Connect

    Rothrock, R.B.

    1998-09-01

    This report summarizes thermal hydraulic analyses performed to support HFIR operation at 100 MW and 468 psig pressure using specially selected fuel elements. The analyses were performed with the HFIR steady state heat transfer code, originally developed during HFIR design. This report addresses the increased core heat removal capability which can be achieved in fuel elements having coolant channel thicknesses that exceed the minimum requirements of the HFIR fuel fabrication specifications. Specific requirements for the minimum value of effective uniform as-built coolant channel thickness are established for fuel elements to be used at 100 MW. The burnout correlation currently used in the steady-state heat transfer code was also compared with more recent experimental results for stability of high-velocity flow in narrow heated channels, and the burnout correlation was found to be conservative with respect to flow stability at typical HFIR hot channel exit conditions at full power.

  14. Single-element coaxial injector for rocket fuel

    NASA Technical Reports Server (NTRS)

    Larson, L. L.

    1969-01-01

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

  15. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

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

  16. 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.

  17. 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.

  18. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

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

  20. 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

  1. Constitutive and heat-inducible heat shock element binding activities of heat shock factor in a group of filamentous fungi

    PubMed Central

    Xavier, Ilungo J.; Khachatourians, George G.; Ovsenek, Nick

    1999-01-01

    This study represents the initial characterization of the heat shock factor (HSF) in filamentous fungi. We demonstrate that HSFs from Beauveria bassiana, Metarhizium anisopliae, Tolypocladium nivea, Paecilomyces farinosus, and Verticillium lecanii bind to the heat shock element (HSE) constitutively (non-shocked), and that heat shock resulted in increased quantities and decreased mobility of HSF-HSE complexes. The monomeric molecular mass of both heat-induced and constitutive HSFs was determined to be 85.8 kDa by UV-crosslinking and the apparent molecular masses of the native HSF-HSE complexes as determined by pore exclusion gradient gel electrophoresis was 260 and 300 kDa, respectively. Proteolytic band clipping assays using trypsin and chymotrypsin revealed an identical partial cleavage profile for constitutive and heat-induced HSF-HSE complexes. Thus, it appears that both constitutive and heat-inducible complexes are formed by trimers composed of the same HSF molecule which undergoes conformational changes during heat shock. The mobility difference between the complexes was not abolished by enzymatic dephosphorylation and deglycosylation, indicating that the reduced mobility of the heat-induced HSF is probably due to a post-translational modification other than phosphorylation or glycosylation. PMID:10590835

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

    PubMed

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

    1985-01-01

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

  3. 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)

  4. 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.

  5. Heat Producing Elements and the Energy Budget of Earth's Core

    NASA Astrophysics Data System (ADS)

    Chidester, B.; Rahman, Z.; Righter, K.; Campbell, A.

    2016-12-01

    The proposed high thermal conductivity of Fe at high pressures and temperatures suggests that the core might require an energy source such as long-lived radioactive decay to explain the presence of a geomagnetic field early in Earth's history. These elements are strongly lithophile at ambient conditions, but may become less so at high P-T conditions. We have measured the metal-silicate partitioning of U and K spanning the high P-T conditions expected for core formation in a magma ocean, using FIB/SEM analyses of laser heated diamond anvil cell experiments. The partitioning of U is strongly dependent on temperature, as well as S in the core and the polymerization of the silicate melt. The partitioning of K is somewhat dependent on the O content of the metal, which is, in turn, dependent on the pressure and temperature conditions in a magma ocean. At core-forming conditions K partitioning is comparable to that of U, but the heat producing effect is significantly lower due to the small fraction of radiogenic K compared to the bulk element. Together, partitioning of U and K during differentiation of the core at an average P-T condition of 55 GPa and 4200 K produce a decrease of 100 K in the inferred initial core-mantle boundary temperature. This temperature difference increases significantly if differentiation were to occur at higher temperatures, similar to that expected after a giant impact, because of increased partitioning of U at those temperatures.

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

    NASA Technical Reports Server (NTRS)

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

    1974-01-01

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

  8. PROCESS OF MAKING FUEL ELEMENTS FOR NEUTRONIC REACTORS AND ARTICLES PRODUCED THEREBY

    DOEpatents

    Bostrom, W.A.; Roof, R.B. Jr.

    1960-04-26

    The novel fuel element is prepared by surrounding a core of U/sub 3/Si with a barrier of various specified alloys or metals, placing a jacket of zirconium around the barrier, and integrally bonding the assembly.

  9. Accuracy of trace element determinations in alternate fuels

    NASA Technical Reports Server (NTRS)

    Greenbauer-Seng, L. A.

    1980-01-01

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

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

    SciTech Connect

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

    1998-06-01

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

  11. 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.

  12. 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.

  13. An investigation of the influence of heating modes on ignition and pyrolysis of woody wildland fuel

    Treesearch

    B.L. Yashwanth; B. Shotorban; S. Mahalingam; D.R. Weise

    2015-01-01

    The ignition of woody wildland fuel modeled as a one-dimensional slab subject to various modes of heating was investigated using a general pyrolysis code, Gpyro. The heating mode was varied by applying different convective and/or radiative, time-dependent heat flux boundary conditions on one end of the slab while keeping the other end insulated. Dry wood properties...

  14. 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 on...

  15. 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 on...

  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 on...

  17. 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..., 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. 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.

  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. Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.

    2017-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). Last year NTREES was successfully used to satisfy a testing milestone for the Nuclear Cryogenic Propulsion Stage (NCPS) project and met or exceeded all required objectives.

  1. 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...

  2. 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...

  3. 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%.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

    DTIC Science & Technology

    1989-05-25

    phase solution. The second corrector applies the pressure equation again with the updated tempera- tures. At the end of the process three estimates of...node model of a fuel particle is perfortned on the solid phase . The model code is applied to a seri s steady state and transient problems, varying...29 CHAPTER 3 THE GAS PHASE

  12. 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.

  13. 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.

  14. 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)

  15. Fuel and Carbon Dioxide Emissions Savings Calculation Methodology for Combined Heat and Power Systems

    EPA Pesticide Factsheets

    This paper provides the EPA Combined Heat and Power Partnership's recommended methodology for calculating fuel and carbon dioxide emissions savings from CHP compared to SHP, which serves as the basis for the EPA's CHP emissions calculator.

  16. On-Line Measurement of Heat of Combustion of Gaseous Hydrocarbon Fuel Mixtures

    NASA Technical Reports Server (NTRS)

    Sprinkle, Danny R.; Chaturvedi, Sushil K.; Kheireddine, Ali

    1996-01-01

    A method for the on-line measurement of the heat of combustion of gaseous hydrocarbon fuel mixtures has been developed and tested. The method involves combustion of a test gas with a measured quantity of air to achieve a preset concentration of oxygen in the combustion products. This method involves using a controller which maintains the fuel (gas) volumetric flow rate at a level consistent with the desired oxygen concentration in the combustion products. The heat of combustion is determined form a known correlation with the fuel flow rate. An on-line computer accesses the fuel flow data and displays the heat of combustion measurement at desired time intervals. This technique appears to be especially applicable for measuring heats of combustion of hydrocarbon mixtures of unknown composition such as natural gas.

  17. Performance testing of refractory alloy-clad fuel elements for space reactors

    SciTech Connect

    Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

    1985-01-01

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO/sub 2/) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts.

  18. Preliminary Heat Transfer Characteristics of RP-2 Fuel as Tested in the High Heat Flux Facility (PREPRINT)

    DTIC Science & Technology

    2005-11-21

    2005. This facility, coupled with established heated tube facilities, such as the NASA Glenn Heated Tube Facility ( HTF ) 3 and the Wright-Patterson...operation differs from the existing facilities in both construction and heating elements. Both the NASA Glenn HTF and the Wright-Patterson Phoenix Rig...ft/sec leading to heat fluxes of up to 100 BTU/in2 sec.1 In similarity to the HTF , the HHFF uses oxygen-free grade copper walled test sections

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

    SciTech Connect

    Pitner, A.L.

    1997-04-14

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

  20. 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.

  1. 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.

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

    NASA Astrophysics Data System (ADS)

    Tondin, Julio Benedito Marin; Filho, Tufic Madi

    2011-08-01

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

  3. Plasma processing of spent nuclear fuel by two-frequency ion cyclotron resonance heating

    SciTech Connect

    Timofeev, A. V.

    2009-11-15

    A previously developed method for analyzing the plasma processing of spent nuclear fuel is generalized to a plasma containing multicharged fuel ions. In such a plasma, ion cyclotron resonance heating of nuclear ash ions should be carried out in two monochromatic RF fields of different frequencies, provided that the fraction of {xi} multicharged ions is small, {xi} {<=} 0.1, a condition that substantially restricts the productivity of systems for processing spent nuclear fuel. Ways of overcoming this difficulty are discussed.

  4. Results of simulated abnormal heating events for full-length nuclear fuel rods

    SciTech Connect

    Guenther, R.J.

    1983-01-01

    Full-length nuclear fuel rods were tested in a furnace to simulate the slow heating rates postulated for commercial pressurized water reactor fuel rods exposed to an overheating event in a storage cask. Fuel rod temperatures and internal gas pressures were monitored during the test and are presented along with mensural data for cladding. Metallography of the cladding provided data on grain growth, hydriding, oxidation, cladding stresses, and the general nature of the failures.

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. Determination of the NPP Kr\\vsko spent fuel decay heat

    NASA Astrophysics Data System (ADS)

    Kromar, Marjan; Kurinčič, Bojan

    2017-07-01

    Nuclear fuel is designed to support fission process in a reactor core. Some of the isotopes, formed during the fission, decay and produce decay heat and radiation. Accurate knowledge of the nuclide inventory producing decay heat is important after reactor shut down, during the fuel storage and subsequent reprocessing or disposal. In this paper possibility to calculate the fuel isotopic composition and determination of the fuel decay heat with the Serpent code is investigated. Serpent is a well-known Monte Carlo code used primarily for the calculation of the neutron transport in a reactor. It has been validated for the burn-up calculations. In the calculation of the fuel decay heat different set of isotopes is important than in the neutron transport case. Comparison with the Origen code is performed to verify that the Serpent is taking into account all isotopes important to assess the fuel decay heat. After the code validation, a sensitivity study is carried out. Influence of several factors such as enrichment, fuel temperature, moderator temperature (density), soluble boron concentration, average power, burnable absorbers, and burnup is analyzed.

  10. 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.

  11. Chemical Gradients in Crud on Boiling Water Reactor Fuel Elements

    SciTech Connect

    D. L. Porter; D. E. Janney

    2007-04-01

    Crud (radioactive corrosion products formed inside nuclear reactors is a major problem in commercial power-producing nuclear reactors. Although there are numerous studies of simulated (non-radioactive) crud, characteristics of crud from actual reactors are rarely studied. This study reports scanning electron microscope (SEM) studies of fragments of crud from a commercially operating boiling water reactor. Chemical analyses in the SEM indicated that the crud closest to the outer surfaces of the fuel pins in some areas had Fe:Zn ratios close to 2:1, which decreased away from the fuel pin in some of the fragments. In combination with transmission electron microsope analyses (published elsewhere), these results suggest that the innermost layer of crud in some areas may consist of franklinite (ZnFe2O4, also called zinc spinel), while outer layers in these areas may be predominantly iron oxides.

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

    DTIC Science & Technology

    1990-10-01

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

  13. Differential recognition of heat shock elements by members of the heat shock transcription factor family.

    PubMed

    Yamamoto, Noritaka; Takemori, Yukiko; Sakurai, Mayumi; Sugiyama, Kazuhisa; Sakurai, Hiroshi

    2009-04-01

    Heat shock transcription factor (HSF), an evolutionarily conserved stress response regulator, forms trimers and binds to heat shock element (HSE), comprising at least three continuous inverted repeats of the sequence 5'-nGAAn-3'. The single HSF of yeast is also able to bind discontinuously arranged nGAAn units. We investigated interactions between three human HSFs and various HSE types in vitro, in yeast cells, and in HeLa cells. Human HSF1, a stress-activated regulator, preferentially bound to continuous HSEs rather than discontinuous HSEs, and heat shock of HeLa cells caused expression of reporter genes containing continuous HSEs. HSF2, whose function is implicated in neuronal specification and spermatogenesis, exhibited a slightly higher binding affinity to discontinuous HSEs than did HSF1. HSF4, a protein required for ocular lens development, efficiently recognized discontinuous HSEs in a trimerization-dependent manner. Among four human gamma-crystallin genes encoding structural proteins of the lens, heat-induced HSF1 preferred HSEs on the gammaA-crystallin and gammaB-crystallin promoters, whereas HSF4 preferred HSE on the gammaC-crystallin promoter. These results suggest that the HSE architecture is an important determinant of which HSF members regulate genes in diverse cellular processes.

  14. Evaluation of Biomass-Derived Distillate Fuel as Renewable Heating Oil

    SciTech Connect

    Mante, Ofei D.; Butcher, Thomas A.; Wei, George; Trojanowski, Rebecca; Sanchez, Vicente

    2015-09-18

    The utilization of advanced biofuels in stationary applications, such as home heating, is considered as an early entry point for biomass-derived fuels into the distillate fuel market sector. Two renewable fuels produced by a biomass fluidized catalytic cracking (BFCC) process, followed by hydroprocessing and fractionation, were tested. The evaluation was performed on a pure (100%) distillate fraction, 50% blend of the distillate fraction with petroleum-based heating oil, and 20% blend of a heavier gas oil fraction. Combustion experiments were carried out in a transparent quartz chamber and a typical oil-fired residential boiler. The flame stability, size, and shape produced by the fuels were examined. The flue gas was analyzed for O2, CO, NOx, and smoke. The elastomer compatibility test was performed with nitrile slabs at 43 °C for 1 month. Fuel stability was examined at 80 °C for 1 week. The results from the combustion studies suggest that the distillate fuel blends could be used as alternative fuels to No. 2 heating oil, even up to 100% without any operational issues. The distillate fuels were found to be stable. and the nitrile slab volume swell (~10%) suggests that the fuel could be compatible to legacy elastomers.

  15. Evaluation of Biomass-Derived Distillate Fuel as Renewable Heating Oil

    DOE PAGES

    Mante, Ofei D.; Butcher, Thomas A.; Wei, George; ...

    2015-09-18

    The utilization of advanced biofuels in stationary applications, such as home heating, is considered as an early entry point for biomass-derived fuels into the distillate fuel market sector. Two renewable fuels produced by a biomass fluidized catalytic cracking (BFCC) process, followed by hydroprocessing and fractionation, were tested. The evaluation was performed on a pure (100%) distillate fraction, 50% blend of the distillate fraction with petroleum-based heating oil, and 20% blend of a heavier gas oil fraction. Combustion experiments were carried out in a transparent quartz chamber and a typical oil-fired residential boiler. The flame stability, size, and shape produced bymore » the fuels were examined. The flue gas was analyzed for O2, CO, NOx, and smoke. The elastomer compatibility test was performed with nitrile slabs at 43 °C for 1 month. Fuel stability was examined at 80 °C for 1 week. The results from the combustion studies suggest that the distillate fuel blends could be used as alternative fuels to No. 2 heating oil, even up to 100% without any operational issues. The distillate fuels were found to be stable. and the nitrile slab volume swell (~10%) suggests that the fuel could be compatible to legacy elastomers.« less

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  1. 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.

  2. 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.

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

    DOEpatents

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ; Heinz, Robert [Ludwigshafen, DE

    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.

  4. 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.

  5. 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.

  6. Salt transport extraction of transuranium elements from lwr fuel

    DOEpatents

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

    1992-01-01

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

  7. 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/.

  8. Accuracy of trace element determinations in alternate fuels

    NASA Technical Reports Server (NTRS)

    Greenbauer-Seng, L. A.

    1980-01-01

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

  9. 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

  10. Apollo 12 Mission image - Alan Bean unloads ALSEP RTG fuel element

    NASA Image and Video Library

    1969-11-19

    AS12-46-6790 (19 Nov. 1969) --- Astronaut Alan L. Bean, lunar module pilot, is photographed at quadrant II of the Lunar Module (LM) during the first Apollo 12 extravehicular activity (EVA) on the moon. This picture was taken by astronaut Charles Conrad Jr., commander. Here, Bean is using a fuel transfer tool to remove the fuel element from the fuel cask mounted on the LM's descent stage. The fuel element was then placed in the Radioisotope Thermoelectric Generator (RTG), the power source for the Apollo Lunar Surface Experiments Package (ALSEP) which was deployed on the moon by the two astronauts. The RTG is next to Bean's right leg. While astronauts Conrad and Bean descended in the LM "Intrepid" to explore the Ocean of Storms region of the moon, astronaut Richard F. Gordon Jr., command module pilot, remained with the Command and Service Modules (CSM) "Yankee Clipper" in lunar orbit.

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

    SciTech Connect

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

    1986-10-01

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

  12. 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].

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

    SciTech Connect

    Sher, R.

    1983-02-01

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

  14. 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)

  15. Commercialisation of fuel cells for combined heat and power (CHP) application

    NASA Astrophysics Data System (ADS)

    Packer, Julian

    1992-01-01

    Combined heat and power or co-generation is an ideal application for the fuel cell. This paper has been written from the perspective of a current designer, builder and operator of small-scale (i.e. sub 1 MW) combined heat and power. Conventional current CHP is described together with typical applications. The perceived advantages of fuel cells are also discussed together with the potential for fuel cells opening up currently unapproachable markets. Various matters relevant to the application of fuel cells are also described including: initial and life costs for fuel cells CHP systems; maintenance requirements, security of supply requirements. In addition to these commercial aspects, technical issues including interfacing to building systems, control, protection, monitoring, operating procedures and performance are also discussed.

  16. 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.

  17. 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.

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

    DOEpatents

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

    2013-09-03

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

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

    DOEpatents

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

    2011-03-01

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

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

    DOEpatents

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

    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.

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

    SciTech Connect

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

    2015-07-01

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

  2. 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.

  3. An advanced method for transient temperature calculation in fuel element structural analysis

    SciTech Connect

    Lassmann, K.; Preusser, T.

    1983-03-01

    An advanced method has been developed for the specific purpose of calculating temperatures in fuel element structural analysis. Fuel, cladding, coolant, and structural temperatures are treated by a single system of equations. Melting of the fuel and cladding and boiling of the coolant are included in the model. The method is compared to other solution techniques. The thermal characteristics of the finite element method (FEM) and finite difference method (FDM) transient calculations are compared. The present method includes FDM and FEM algorithms as special cases; an optimum combination of both techniques is the standard usage. Explicit, implicit, or Crank-Nicholson integration procedures are possible. The method is fast running, reliable, and has no stability problems. The new method has been implemented into the temperature calculation subcode system TEMPER for use with URANUS or other fuel element codes. Special attention has been given to user requirements (e.g., an automatic time-step control). The URANUS code, with this subcode system TEMPER, has been applied successfully to difficult fast breeder fuel rod analysis including transient overpower, loss of flow, local coolant blockage, and specific carbide fuel experiments.

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

    NASA Technical Reports Server (NTRS)

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

    1973-01-01

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

  5. Effect of heat rate constraint on minimum-fuel synergetic plane change

    NASA Technical Reports Server (NTRS)

    Mease, Kenneth D.; Utashima, Masayoshi

    1991-01-01

    The synergetic plane change offers substantial fuel savings over the pure-propulsive alternative for certain noncoplanar orbital transfers. On the other hand, the thermal environment for a synergetic plane change vehicle can be quite severe. The minimum-fuel controls are computed approximately by parametrizing the controls and solving the resulting nonlinear programming problem. By considering several different levels of heat rate constraint, we characterize how the control strategy should be modified in order to keep the heat rate below the specified limit. Flight on the heat rate constraint boundary at high angle of attack is the key characteristic.

  6. 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.

  7. 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

  8. 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.

  9. 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.

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

    NASA Astrophysics Data System (ADS)

    Newman, Chris; Hansen, Glen; Gaston, Derek

    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 a subset of phenomena that are important in the development of an analysis capability for fuel performance calculations, focusing on thermomechanics and diffusion within UO 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

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

    SciTech Connect

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

    1994-10-01

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

  17. 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.

  18. Liquid fueled external heating system for STM4-120 Stirling engine

    NASA Astrophysics Data System (ADS)

    Meijer, R. J.; Ziph, B.; Godett, T. M.

    1985-12-01

    The STM4-120 Stirling engine, currently under development at Stirling Thermal Motors, Inc., is a 40 kW variable stroke engine with indirect heating using a sodium heat pipe. The engine is functionally separated into an application independent Energy Conversion Unit (ECU) consisting of the Stirling cycle and drive heated by condensing sodium and the application dependent External Heating System (EHS), designed to supply the ECU with sodium vapor heated by the particular energy source, connected by tubes with mechanical couplings. This paper describes an External Heating System for the STM4-120 ECU designed for the combustion of liquid fuel, comprised of a recuperative preheater, a combustion chamber, and a heat exchanger/evaporator where heat is transferred from the flue gas to the sodium causing it to evaporate. The design concept and projected performance are described and discussed.

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

    NASA Astrophysics Data System (ADS)

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

    1997-01-01

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

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

    DOEpatents

    Duckworth, W.H.

    1957-12-01

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

  1. Heating with Solid Fuels. A Unit of Instruction.

    ERIC Educational Resources Information Center

    Rockel, Edward

    Designed for use in industrial education programs at the secondary school level, this unit focuses on residential space heating although applications can be made to commercial settings. Wood heat is emphasized but coal-fired appliances and other energy sources are considered. Educational objectives with instructional strategies are provided for…

  2. Heating with Solid Fuels. A Unit of Instruction.

    ERIC Educational Resources Information Center

    Rockel, Edward

    Designed for use in industrial education programs at the secondary school level, this unit focuses on residential space heating although applications can be made to commercial settings. Wood heat is emphasized but coal-fired appliances and other energy sources are considered. Educational objectives with instructional strategies are provided for…

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. Experimental studies of heat exchange for sodium boiling in the fuel assembly model: Safety substantiation of a promising fast reactor

    NASA Astrophysics Data System (ADS)

    Khafizov, R. R.; Poplavskii, V. M.; Rachkov, V. I.; Sorokin, A. P.; Trufanov, A. A.; Ashurko, Yu. M.; Volkov, A. V.; Ivanov, E. F.; Privezentsev, V. V.

    2017-01-01

    Numerical simulation of the ULOF-type accident development in a fast reactor with sodium coolant performed using the COREMELT code indicates that sodium boiling in the active core takes place. The boiling is accompanied by oscillations of the technological parameters of the reactor installation; these oscillations can go on during several tens of seconds. In this case, it is possible that a stable regime of removal of heat from residual energy release is implemented. The model of the two-phase coolant flow applied in the code has an important effect on the numerical results; that is why this model needs experimental verification. For eliminating the development of an accident resulting in destruction of the active core elements, a structural solution is proposed; the essence of it is the application of the sodium void above the reactor active core. The experimental installation was developed and the heat exchange at sodium boiling in the model fuel assembly of the fast reactor in the regimes of natural and forced circulation in the presence of the sodium void and the top end shield was studied. It was demonstrated that, in the presence of the sodium void, it is possible to provide long-term cooling of the fuel assembly for a thermal flux density on the fuel element simulator surface of up to 140 and 170 kW/m2 in the natural and forced circulation modes, respectively. The obtained data are used for more precise determination of the numerical model of sodium boiling in the fuel assembly and verification of the COREMELT code.

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

    PubMed

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

    2015-01-01

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

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

    PubMed Central

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

    2015-01-01

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

  11. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

    SciTech Connect

    Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; Hunn, John D.; Reber, Edward L.

    2016-05-18

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (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% fissions per initial metal atom (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 failed 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. As a result, 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

  12. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

    SciTech Connect

    Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; Hunn, John D.; Reber, Edward L.

    2016-05-18

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (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% fissions per initial metal atom (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 failed 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. As a result, 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

  13. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

    DOE PAGES

    Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; ...

    2016-05-18

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (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% fissions per initial metal atom (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,more » 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 failed 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. As a result, 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

  14. 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.

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

    NASA Technical Reports Server (NTRS)

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

    1987-01-01

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

  16. 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.

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

    SciTech Connect

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

    1980-01-01

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

  18. 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.

  19. 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.

  20. Demand for waste as fuel in the swedish district heating sector: a production function approach.

    PubMed

    Furtenback, Orjan

    2009-01-01

    This paper evaluates inter-fuel substitution in the Swedish district heating industry by analyzing almost all the district heating plants in Sweden in the period 1989-2003, specifically those plants incinerating waste. A multi-output plant-specific production function is estimated using panel data methods. A procedure for weighting the elasticities of factor demand to produce a single matrix for the whole industry is introduced. The price of waste is assumed to increase in response to the energy and CO2 tax on waste-to-energy incineration that was introduced in Sweden on 1 July 2006. Analysis of the plants involved in waste incineration indicates that an increase in the net price of waste by 10% is likely to reduce the demand for waste by 4.2%, and increase the demand for bio-fuels, fossil fuels, other fuels and electricity by 5.5%, 6.0%, 6.0% and 6.0%, respectively.

  1. Preliminary Heat Transfer Characteristics of RP-2 Fuel as Tested in the High Heat Flux Facility (POSTPRINT)

    DTIC Science & Technology

    2005-11-21

    operational in October 2005. This facility, coupled with established heated tube facilities, such as the NASA Glenn Heated Tube Facility ( HTF ) 3 and...Facility design and operation differs from the existing facilities in both construction and heating elements. Both the NASA Glenn HTF and the Wright...flow rates of up to 450 ft/sec leading to heat fluxes of up to 100 BTU/in2 sec.1 In similarity to the HTF , the HHFF uses oxygen-free grade copper

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

    NASA Astrophysics Data System (ADS)

    Govers, K.; Verwerft, M.

    2016-09-01

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

  3. 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.

  4. Airflow Obstruction and Use of Solid Fuels for Cooking or Heating: BOLD Results.

    PubMed

    Amaral, André F S; Patel, Jaymini; Kato, Bernet S; Obaseki, Daniel O; Lawin, Hervé; Tan, Wan C; Juvekar, Sanjay K; Harrabi, Imed; Studnicka, Michael; Wouters, Emiel F M; Loh, Li-Cher; Bateman, Eric D; Mortimer, Kevin; Buist, A Sonia; Burney, Peter G J

    2017-09-12

    Evidence supporting the association of COPD or airflow obstruction with use of solid fuels is conflicting and inconsistent. To assess the association of airflow obstruction with self-reported use of solid fuels for cooking or heating. We analysed 18,554 adults from the BOLD study, who had provided acceptable post-bronchodilator spirometry measurements and information on use of solid fuels. The association of airflow obstruction with use of solid fuels for cooking or heating was assessed by sex, within each site, using regression analysis. Estimates were stratified by national income and meta-analysed. We carried out similar analyses for spirometric restriction, chronic cough and chronic phlegm. We found no association between airflow obstruction and use of solid fuels for cooking or heating (ORmen=1.20, 95%CI 0.94-1.53; ORwomen=0.88, 95%CI 0.67-1.15). This was true for low/middle and high income sites. Among never smokers there was also no evidence of an association of airflow obstruction with use of solid fuels (ORmen=1.00, 95%CI 0.57-1.76; ORwomen=1.00, 95%CI 0.76-1.32). Overall, we found no association of spirometric restriction, chronic cough or chronic phlegm with the use of solid fuels. However, we found that chronic phlegm was more likely to be reported among female never smokers and those who had been exposed for ≥20 years. Airflow obstruction assessed from post-bronchodilator spirometry was not associated with use of solid fuels for cooking or heating.

  5. 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.

  6. NASA Data Helps Track Heat Potential Fueling Rita

    NASA Image and Video Library

    2005-09-26

    Tropical Cyclone Heat Potential TCHP field in the Gulf of Mexico during September 22, 2005. The path of Hurricane Rita is indicated with circles spaced every 3 hours with their size and color representing intensity see legend.

  7. Two dimensional finite element heat transfer models for softwood

    Treesearch

    Hongmei Gu; John F. Hunt

    2004-01-01

    The anisotropy of wood creates a complex problem for solving heat and mass transfer problems that require analyses be based on fundamental material properties of the wood structure. Most heat transfer models use average thermal properties across either the radial or tangential directions and have not differentiated the effects of cellular alignment, earlywood/latewood...

  8. 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.

  9. 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.

  10. Economic analysis of a combined heat and power molten carbonate fuel cell system

    NASA Astrophysics Data System (ADS)

    Hengeveld, Derek W.; Revankar, Shripad T.

    Fuel cells can be attractive for use as stationary combined heat and power (CHP) systems. Molten carbonate fuel cell (MCFC) power plants are prime candidates for the utilization of fossil based fuels to generate high efficiency ultra clean power. However, fuel cells are considerably more expensive than comparable conventional technologies and therefore a careful analysis of the economics must be taken. This work presents analysis on the feasibility of installing both a FuelCell Energy DFC ® 1500MA and 300MA system for use at Adams Thermal Systems, a manufacturing facility in the U.S. Midwest. The paper examined thoroughly the economics driving the appropriateness of this measure. In addition, a parametric study was conducted to determine scenarios including variation in electric and natural gas rates along with reduced installation costs.

  11. 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.

  12. Pyrochemical recovery of actinide elements from spent light water reactor fuel

    SciTech Connect

    Johnson, G.K.; Pierce, R.D.; Poa, D.S.; McPheeters, C.C.

    1994-01-01

    Argonne National Laboratory is investigating salt transport and lithium pyrochemical processes for recovery of transuranic (TRU) elements from spent light water reactor fuel. The two processes are designed to recover the TRU elements in a form compatible with the Integral Fast Reactor (IFR) fuel cycle. The IFR is uniquely effective in consuming these long-lived TRU elements. The salt transport process uses calcium dissolved in Cu-35 wt % Mg in the presence of a CaCl{sub 2} salt to reduce the oxide fuel. The reduced TRU elements are separated from uranium and most of the fission products by using a MgCl{sub 2} transport salt. The lithium process, which does not employ a solvent metal, uses lithium in the presence of a LiCl salt as the reductant. After separation from the salt, the reduced metal is introduced into an electrorefiner, which separates the TRU elements from the uranium and fission products. In both processes, reductant and reduction salt are recovered by electrochemical decomposition of the oxide reaction product.

  13. Process integration methodology for natural gas-fueled heat pumps and cogeneration systems

    NASA Astrophysics Data System (ADS)

    Rossiter, Alan P.

    1988-11-01

    A process integration methodology was developed for analyzing industrial processes, identifying those that will benefit from natural gas fueled heat pumps and cogeneration system as well as novel, process-specific opportunities for further equipment improvements, including performance targets. The development included the writing of software to assist in implementing the methodology and application of the procedures in studies using both literature data and plant operating data. These highlighted potential applications for gas fueled heat pumps in ethylene processes and liquor distilling plants, and slightly less attractive opportunities in a number of other plants. Many of the processes studied showed excellent potentials for cogeneration applications.

  14. Design options for achieving a rapidly variable heat-to-power ratio in a combined heat and power (CHP) fuel cell system (FCS)

    NASA Astrophysics Data System (ADS)

    Colella, Whitney

    This article calls for a change in paradigm within the fuel cells industry such that it focuses less on solely maximizing a fuel cell's electrical efficiency, and more on a fuel cell system's (FCS) overall combined thermal and electrical efficiency, as defined in relation to the instantaneous demand for heat and electricity. Based on market needs in the power generation sector, it emphasizes the need to develop FCSs such that they can achieve a heat-to-power ratio that can be rapidly varied. This article then delineates engineering methods to achieve a rapidly variable heat-to-power ratio for a combined heat and power (CHP) FCS.

  15. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2005-05-01

    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.

  17. Binary effect of fly ash and palm oil fuel ash on heat of hydration aerated concrete.

    PubMed

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

    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.

  18. 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

  19. 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.

  20. Test design description Volume 2, Part 1. IFR-1 metal fuel irradiation test (AK-181) element as-built data

    SciTech Connect

    Dodds, N. E.

    1986-06-01

    The IFR-1 Test, designated as the AK-181 Test Assembly, will be the first irradiation test of wire wrapped, sodium-bonded metallic fuel elements in the Fast Flux Test Facility (FFTF). The test is part of the Integral Fast Reactor (IFR) fuels program conducted by Argonne National Laboratory (ANL) in support of the Innovative Reactor Concepts Program sponsored by the US Department of Energy (DOE). One subassembly, containing 169 fuel elements, will be irradiated for 600 full power days to achieve 10 at.% burnup. Three metal fuel alloys (U-10Zr, U-8Pu-10Zr) will be irradiated in D9 cladding tubes. The metal fuel elements have a fuel-smeared density of 75% and each contains five slugs. The enriched zone contains three slugs and is 36-in. long. One 6.5-in. long depleted uranium axial blanket slug (DU-10Zr) was loaded at each end of the enriched zone. the fuel elements were fabricated at ANL-W and delivered to Westinghouse-Hanford for wirewrapping and assembly into the test article. This Test Design Description contains relevant data on compositions, densities, dimensions and weights for the cast fuel slugs and completed fuel elements. The elements conform to the requirements in MG-22, "Users` Guide for the Irradiation of Experiments in the FTR."

  1. 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.

  2. 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.

  3. 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.

  4. Genomic Heat Shock Element Sequences Drive Cooperative Human Heat Shock Factor 1 DNA Binding and Selectivity*

    PubMed Central

    Jaeger, Alex M.; Makley, Leah N.; Gestwicki, Jason E.; Thiele, Dennis J.

    2014-01-01

    The heat shock transcription factor 1 (HSF1) activates expression of a variety of genes involved in cell survival, including protein chaperones, the protein degradation machinery, anti-apoptotic proteins, and transcription factors. Although HSF1 activation has been linked to amelioration of neurodegenerative disease, cancer cells exhibit a dependence on HSF1 for survival. Indeed, HSF1 drives a program of gene expression in cancer cells that is distinct from that activated in response to proteotoxic stress, and HSF1 DNA binding activity is elevated in cycling cells as compared with arrested cells. Active HSF1 homotrimerizes and binds to a DNA sequence consisting of inverted repeats of the pentameric sequence nGAAn, known as heat shock elements (HSEs). Recent comprehensive ChIP-seq experiments demonstrated that the architecture of HSEs is very diverse in the human genome, with deviations from the consensus sequence in the spacing, orientation, and extent of HSE repeats that could influence HSF1 DNA binding efficacy and the kinetics and magnitude of target gene expression. To understand the mechanisms that dictate binding specificity, HSF1 was purified as either a monomer or trimer and used to evaluate DNA-binding site preferences in vitro using fluorescence polarization and thermal denaturation profiling. These results were compared with quantitative chromatin immunoprecipitation assays in vivo. We demonstrate a role for specific orientations of extended HSE sequences in driving preferential HSF1 DNA binding to target loci in vivo. These studies provide a biochemical basis for understanding differential HSF1 target gene recognition and transcription in neurodegenerative disease and in cancer. PMID:25204655

  5. 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.

  6. 40 CFR Table 1 to Subpart Ja of... - Molar Exhaust Volumes and Molar Heat Content of Fuel Gas Constituents

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 7 2014-07-01 2014-07-01 false Molar Exhaust Volumes and Molar Heat... Exhaust Volumes and Molar Heat Content of Fuel Gas Constituents Constituent MEVa dscf/mol MHCb Btu/mol... standard conditions of 68 °F and 1 atmosphere. b MHC = molar heat content (higher heating value basis), Btu...

  7. 40 CFR Table 1 to Subpart Ja of... - Molar Exhaust Volumes and Molar Heat Content of Fuel Gas Constituents

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 7 2013-07-01 2013-07-01 false Molar Exhaust Volumes and Molar Heat... Exhaust Volumes and Molar Heat Content of Fuel Gas Constituents Constituent MEVa dscf/mol MHCb Btu/mol... standard conditions of 68 °F and 1 atmosphere. b MHC = molar heat content (higher heating value basis), Btu...

  8. Fuel Composition Analysis of Endothermically Heated JP-8 Fuel for Use in a Pulse Detonation Engine

    DTIC Science & Technology

    2008-06-01

    Cengel , 2006:13). Total amount of heat transfer (Q ) is expressed in Eq. 33: 65 ( )pQ Q t t mc T= Δ = Δ Δ...A.1) where is the mass flow and cp is the constant pressure specific heat ( Cengel , 2006:458). The mean... Cengel , 2006:510). The convection heat transfer coefficient (h) was found by Eq. A.2: kh N D = u

  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. 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.

  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. New Insights Into the Heat Sources of Mantle Plumes, or: Where Does all the Heat Come From, Heat Producing Elements, Advective or Conductive Heat Flow?

    NASA Astrophysics Data System (ADS)

    Rushmer, T.; Beier, C.; Turner, S.

    2007-12-01

    Melting anomalies in the Earth's upper mantle have often been attributed to the presence of mantle plumes that may originate in the lower mantle, possibly from the core-mantle boundary. Globally, mantle plumes exhibit a large range in buoyancy flux that which is proportional to their temperature and volume. Plumes with higher buoyancy fluxes should have higher temperatures and experience higher degrees of partial melting. Excess heat in mantle plumes could reflect either a) an enrichment of the heat producing elements (HPE: U, Th, K) in their mantle source leading to an increase of heat production by radioactive decay or b) advective or conductive heat transport across the core-mantle boundary. The advective transport of heat may result in a physical contribution of material from the core to the lower mantle. If core material is incorporated into the lower mantle, mantle plumes with a higher buoyancy flux should have higher core tracers, e.g. increased 186Os and Fe concentrations. Geophysical and dynamic modelling indicate that at least Afar, Easter, Hawaii, Louisville and Samoa may all originate at the core-mantle boundary. These plumes encompass the whole range of known buoyancy fluxes from 1.2 Mgs -1(Afar) to 6.5 Mgs -1 (Hawaii) providing evidence that the buoyancy flux is largely independent of other geophysical parameters. In an effort to explore whether the heat producing elements are the cause of excess heat we looked for correlations between fractionation corrected concentrations of the HPE and buoyancy flux. Our results suggest that there is no correlation between HPE concentrations and buoyancy flux (with and without an additional correction for variable degrees of partial melting). As anticipated, K, Th and U are positively correlated with each other (e.g. Hawaii, Iceland and Galapagos have significantly lower concentrations than e.g. Tristan da Cunha, the Canary Islands and the Azores). We also find no correlation between currently available Fe

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. Experimental study of pine forest fuel layer ignition by the steel heated particle

    NASA Astrophysics Data System (ADS)

    Baranovskiy, Nikolay V.; Zakharevich, Arkadiy V.; Osotova, Diana S.

    2015-01-01

    Experimental research results of ignition processes of typical and widespread forest fuel (pine needles) by single heated up to high temperatures metal particles of the cylindrical form are presented. Ignition delays for particles of the various sizes and initial temperatures are resulted. Ignition conditions are established and a number of features of the investigated process mechanism are marked.

  18. 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

  19. 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

  20. Synergistic effects of toxic elements on heat shock proteins.

    PubMed

    Mahmood, Khalid; Jadoon, Saima; 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.