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Sample records for lmfbr fuel subassemblies

  1. Fuel or irradiation subassembly

    DOEpatents

    Seim, O.S.; Hutter, E.

    1975-12-23

    A subassembly for use in a nuclear reactor is described which incorporates a loose bundle of fuel or irradiation pins enclosed within an inner tube which in turn is enclosed within an outer coolant tube and includes a locking comb consisting of a head extending through one side of the inner sleeve and a plurality of teeth which extend through the other side of the inner sleeve while engaging annular undercut portions in the bottom portion of the fuel or irradiation pins to prevent movement of the pins.

  2. Vibrating fuel grapple. [LMFBR

    DOEpatents

    Chertock, A.J.; Fox, J.N.; Weissinger, R.B.

    A reactor refueling method is described which utilizes a vibrating fuel grapple for removing spent fuel assemblies from a reactor core. It incorporates a pneumatic vibrator in the grapple head which allows additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

  3. Fuel cell subassemblies incorporating subgasketed thrifted membranes

    DOEpatents

    Iverson, Eric J.; Pierpont, Daniel M.; Yandrasits, Michael A.; Hamrock, Steven J.; Obradovich, Stephan J.; Peterson, Donald G.

    2013-03-01

    A fuel cell roll good subassembly is described that includes a plurality of individual electrolyte membranes. One or more first subgaskets are attached to the individual electrolyte membranes. Each of the first subgaskets has at least one aperture and the first subgaskets are arranged so the center regions of the individual electrolyte membranes are exposed through the apertures of the first subgaskets. A second subgasket comprises a web having a plurality of apertures. The second subgasket web is attached to the one or more first subgaskets so the center regions of the individual electrolyte membranes are exposed through the apertures of the second subgasket web. The second subgasket web may have little or no adhesive on the subgasket surface facing the electrolyte membrane.

  4. Fuel cell subassemblies incorporating subgasketed thrifted membranes

    DOEpatents

    Iverson, Eric J; Pierpont, Daniel M; Yandrasits, Michael A; Hamrock, Steven J; Obradovich, Stephan J; Peterson, Donald G

    2014-01-28

    A fuel cell roll good subassembly is described that includes a plurality of individual electrolyte membranes. One or more first subgaskets are attached to the individual electrolyte membranes. Each of the first subgaskets has at least one aperture and the first subgaskets are arranged so the center regions of the individual electrolyte membranes are exposed through the apertures of the first subgaskets. A second subgasket comprises a web having a plurality of apertures. The second subgasket web is attached to the one or more first subgaskets so the center regions of the individual electrolyte membranes are exposed through the apertures of the second subgasket web. The second subgasket web may have little or no adhesive on the subgasket surface facing the electrolyte membrane.

  5. Fuel subassembly leak test chamber for a nuclear reactor

    DOEpatents

    Divona, Charles J.

    1978-04-04

    A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained.

  6. Fuels and materials for LMFBR core components

    SciTech Connect

    Cox, C M; Jackson, R J; Straalsund, J L

    1984-04-01

    This paper reviews development of fuels and materials for Liquid Metal Fast Breeder Reactor. Included are the status of fuels and materials technology for LMFBR core components. The fuel assembly for the Fast Flux Test Facility, or FFTF, in operation near Richland, Washington, is described. The outer part of the 12-ft long assembly is called a flow channel or duct. Inside are 217 fuel pins, each containing mixed uranium-plutonium oxide fuel pellets. The comparable schematic for control rod or absorber assembly is also shown. The FFTF absorber assembly contains 61 control rods containing boron carbide pellets. Because FFTF is a test reactor, it does not contain blanket assemblies; however, the Clinch River Breeder Reactor blanket assemblies look very similar to the FFTF fuel assembly, except that they each contain 61 UO/sub 2/ rods. Sizes of various LMFBR fuel assemblies are compared. The Clinch River Breeder Reactor fuel assembly is nearly identical to that of FFTF, except for an increased length to accommodate UO/sub 2/ axial blankets within the fuel pins. The DP-1 design is for a large breeder reactor and uses larger ducts and more fuel pins per assembly. By comparison, the fuel assemblies for EBR-II are much smaller, as is the EBR-II core.

  7. Laser cutting apparatus for nuclear core fuel subassembly

    DOEpatents

    Walch, Allan P.; Caruolo, Antonio B.

    1982-02-23

    The object of the invention is to provide a system and apparatus which employs laser cutting to disassemble a nuclear core fuel subassembly. The apparatus includes a gantry frame (C) which straddles the core fuel subassembly (14), an x-carriage (22) travelling longitudinally above the frame which carries a focus head assembly (D) having a vertically moving carriage (46) and a laterally moving carriage (52), a system of laser beam transferring and focusing mirrors carried by the x-carriage and focusing head assembly, and a shroud follower (F) and longitudinal follower (G) for following the shape of shroud (14) to maintain a beam focal point (44) fixed upon the shroud surface for accurate cutting.

  8. Nonlinear, inelastic fast reactor subassembly interaction analyses

    SciTech Connect

    Sutherland, W.H.; Bard, F.E.

    1983-01-01

    Liquid Metal Fast Breeder Reactor (LMFBR) core structural design is complicated by the trade-offs associated with keeping the subassemblies closely packed for the neutronic considerations and accommodating the volumetric changes associated with irradiation swelling. The environmental variation across the reactor core results in temperature and neutron flux gradients across the subassemblies which in turn cause the subassemblies to bow as well as dilate and grow volumetrically. These deformations in a tightly packed reactor core cause the subassemblies to interact and can potentially result in excessive withdrawal loads during the refueling operations. ABADAN, a general purpose, nonlinear, inelastic, multi-dimensional finite element structural analysis computer code, was developed for the express purpose of solving large nonlinear problems as typified by the above interaction problems. For the subassembly interaction problem ABADAN has been applied to the solution of an interacting radial row of Fast Flux Test Facility (FFTF) fuel assemblies.

  9. Preliminary study: isotopic safeguards techniques (IST) LMFBR fuel cycles

    SciTech Connect

    Persiani, P. J.; Kroc, T. K.

    1980-06-01

    This memorandum presents the preliminary results of the effort to investigate the applicability of isotope correlation techniques (ICT), formulated for the LWR system, to the LMFBR fuel cycle. The detailed isotopic compositional changes with burnup developed for the CRBR was utilized as the reference case. This differs from the usual LMFBR design studies in that the core uranium is natural uranium rather than depleted. Nevertheless, the general isotopic behavior should not differ significantly and does allow an initial insight into the expected behavior of isotopic correlations for the LMFBR power systems such as: the U.K. PFR and reprocessing plant; the French Phenix and Superphenix; and the US reference conceptual design studies (CDS) of homogeneous and heterogeneous LMFBR systems as they are developed.

  10. Heat-transfer calculations for a potted (solid matrix embedded) subassembly

    SciTech Connect

    Betten, P.R.

    1984-01-01

    Standard Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies used in the Experimental Breeder Reactor II (EBR-II) have been investigated for fuel-bundle distortion using a destructive examination method known as potting. The potting technique embeds and permanently fixes the fuel elements in a solid matrix that can be sectioned and polished to reveal details in the internal structure of the elements or subassembly. Thus, an advantage of the potting technique is that it permits investigation of the internal structure of the subassembly in situ, as this structure would be lost or significantly altered during subassembly disassembly. However, since the elements in the subassembly are radioactive, the potting material must efficiently conduct radioactive decay heat to the environment so that the melting or softening temperatures of the potting material are not exceeded. The purpose of this paper is to present the heat transfer calculations for a potted subassembly and to recommend a simplified method for solving similar problems.

  11. Performance of breached LMFBR fuel pins during continued operation

    SciTech Connect

    Lambert, J.D.B.; Strain, R.V.; Gross, K.C.; Hofman, G.L.; Colburn, R.P.; Adamson, M.G.; Ukai, S.

    1985-01-01

    Four EBR-II tests were used to scope the behavior of breached mixed-oxide pins. After release of stored fission gas, delayed-neutron signals were large and easily detected, although not readily correlated with exposed fuel area. No problems were met during reactor operation or fuel handling. Fuel-sodium reaction caused only narrow breaches which released minute amounts of fuel and fission products; the reaction product appeared dense and non-friable. These initial results indicated LMFBR oxide pins could have considerable potential for operating in the breached mode.

  12. Evaluation of copper for divider subassembly in MCO Mark IA and Mark IV scrap fuel baskets

    SciTech Connect

    Graves, C.E.

    1997-09-29

    The K Basin Spent Nuclear Fuel (SNF) Project Multi-Canister Overpack (MCO) subprojection eludes the design and fabrication of a canister that will be used to confine, contain, and maintain fuel in a critically safe array to enable its removal from the K Basins, vacuum drying, transport, staging, hot conditioning, and interim storage (Goldinann 1997). Each MCO consists of a shell, shield plug, fuel baskets (Mark IA or Mark IV), and other incidental equipment. The Mark IA intact and scrap fuel baskets are a safety class item for criticality control and components necessary for criticality control will be constructed from 304L stainless steel. It is proposed that a copper divider subassembly be used in both Mark IA and Mark IV scrap baskets to increase the safety basis margin during cold vacuum drying. The use of copper would increase the heat conducted away from hot areas in the baskets out to the wall of the MCO by both radiative and conductive heat transfer means. Thus copper subassembly will likely be a safety significant component of the scrap fuel baskets. This report examines the structural, cost and corrosion consequences associated with using a copper subassembly in the stainless steel MCO scrap fuel baskets.

  13. FUEL SUBASSEMBLY CONSTRUCTION FOR RADIAL FLOW IN A NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1962-12-25

    An assembly of fuel elements for a boiling water reactor arranged for radial flow of the coolant is described. The ingress for the coolant is through a central header tube, perforated with parallel circumferertial rows of openings each having a lip to direct the coolant flow downward. Around the central tube there are a number of equally spaced concentric trays, closely fitiing the central header tube. Cylindrical fuel elements are placed in a regular pattern around the central tube, piercing the trays. A larger tube encloses the arrangement, with space provided for upward flow of coolart beyond the edge of the trays. (AEC)

  14. Device for gripping and detaching a top nozzle subassembly from a reconstitutable fuel assembly

    SciTech Connect

    Wilson, J.F.; Gjersten, R.K.

    1987-03-03

    This patent describes a reconstitutable fuel assembly including a top nozzle subassembly and guide thimbles. The top nozzle subassembly has a lower adapter plate, hold-down springs and an upper hold-down plate with coolant flow openings defined therethrough. The guide thimbles have upper end portions slidably mounting the lower adapter plate and upper hold-down plate for movement therealong between lower and upper limits. A device is described for gripping and detaching the top nozzle subassembly from the guide thimble upper end portions, comprising: (a) a central spider disposable in overlying relation to the upper hold-down plate; (b) locating lugs disposed radially outwardly from the spider and arranged for alignment with and insertion into the plurality of coolant flow openings in the upper hold-down plate. Each of the locating lugs has an elongated central bore; (c) collars interconnected to the spider, each collar connected to one of the locating lugs and bearing on the hold-down plate when the locating lug is inserted in its respective flow opening; and (d) elongated members received through and rotatable within the respective central bores of the locating lugs.

  15. Analysis of Pu-Only Partitioning Strategies in LMFBR Fuel Cycles

    SciTech Connect

    Samuel Bays; Gilles Youinou

    2013-02-01

    Sodium cooled Fast Reactors (SFR) have been under consideration for production of electricity, fissile material production, and for destruction of transuranics for decades. The neutron economy of a SFR can be operated in one of two ways. One possibility is to operate the reactor in a transuranic burner mode which has been the focus of active R&D in the last 15 years. However, prior to that the focus was on breeding transuranics. This later mode of managing the neutron economy relies on ensuring the maximum fuel utilization possible in such a way as to maximize the amount of plutonium produced per unit of fission energy in the reactor core. The goal of maximizing plutonium production in this study is as fissile feed stock for the production of MOX fuel to be used in Light Water Reactors (LWR). Throughout the l970’s, this fuel cycle scenario was the focus of much research by the Atomic Energy Commission in the event that uranium supplies would be scarce. To date, there has been sufficient uranium to supply the once through nuclear fuel cycle. However, interest in a synergistic relationship Liquid Metal Fast Breeder Reactors (LMFBR) and a consumer LWR fleet persists, prompting this study. This study considered LMFBR concepts with varying additions of axial and radial reflectors. Three scenarios were considered in collaboration with a companion study on the LWR-MOX designs based on plutonium nuclide vectors produced by this study. The first scenario is a LMFBR providing fissile material to make MOX fuel where the MOX part of the fuel cycle is operated in a once-through-then-out mode. The second scenario is the same as the first but with the MOX part of the fuel cycle multi-recycling its own plutonium with LMFBR being used for the make-up feed. In these first two scenarios, plutonium partitioning from the minor actinides (MA) was assumed. Also, the plutonium management strategy of the LMFBR ensured that only the high fissile purity plutonium bred from blankets was

  16. LMFBR fuel assembly design for HCDA fuel dispersal

    DOEpatents

    Lacko, Robert E.; Tilbrook, Roger W.

    1984-01-01

    A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

  17. Irradiation subassembly

    DOEpatents

    Seim, O.S.; Filewicz, E.C.; Hutter, E.

    1973-10-23

    An irradiation subassembly for use in a nuclear reactor is described which includes a bundle of slender elongated irradiation -capsules or fuel elements enclosed by a coolant tube and having yieldable retaining liner between the irradiation capsules and the coolant tube. For a hexagonal bundle surrounded by a hexagonal tube the yieldable retaining liner may consist either of six segments corresponding to the six sides of the tube or three angular segments each corresponding in two adjacent sides of the tube. The sides of adjacent segments abut and are so cut that metal-tometal contact is retained when the volume enclosed by the retaining liner is varied and Springs are provided for urging the segments toward the center of the tube to hold the capsules in a closely packed configuration. (Official Gazette)

  18. Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

    SciTech Connect

    Sonat Sen; Gilles Youinou

    2013-02-01

    It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this case the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)

  19. Bundle duct interaction studies for fuel assemblies. [LMFBR

    SciTech Connect

    Hsia, H.T.S.; Kaplan, S.

    1981-06-01

    It is known that the wire-wrapped rods and duct in an LMFBR are undergoing a gradual structural distortion from the initially uniform geometry under the combined effects of thermal expansion and irradiation induced swelling and creep. These deformations have a significant effect on flow characteristics, thus causing changes in thermal behavior such as cladding temperature and temperature distribution within a bundle. The temperature distribution may further enhance or retard irradiation induced deformation of the bundle. This report summarizes the results of the continuing effort in investigating the bundle-duct interaction, focusing on the need for the large development plant.

  20. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    SciTech Connect

    Riley, D.R.; Dickson, P.W.

    1981-01-01

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle.

  1. Advanced LMFBR fuel cladding susceptability to stress corrosion due to reprocessing impurities

    SciTech Connect

    Henslee, S.P.

    1987-03-01

    The potential degradation of LMFBR fuel cladding alloys by chlorides, when used in metallic fuel systems, was evaluated. The alloys tested were D-9 and HT-9 stainless steels, austenitic and ferritic alloys respectively. These two alloys were tested in parallel with and their performance compared to the austenitic stainless steel Type 316. All alloys were tested for 7400 hours in a stress rupture environment with chloride exposure at either 550/degree/C 650/degree/C. None of the alloys tested were found to exhibit any degradation in time-to-rupture by the presence of chlorides under the conditions imposed during testing. 8 refs., 4 figs., 2 tabs.

  2. Radial plutonium redistribution in mixed-oxide fuel. [LMFBR

    SciTech Connect

    Lawrence, L.A.; Schwinkendorf, K.N.; Karnesky, R.A.

    1981-10-01

    Alpha autoradiographs from all HEDL fuel pin metallography samples are evaluated and catalogued according to different plutonium distribution patterns. The data base is analyzed for effects of fabrication and operating parameters on redistribution.

  3. Two-dimensional computational modeling of sodium boiling in simulated LMFBR fuel-pin bundles

    SciTech Connect

    Dearing, J.F.

    1981-01-01

    Extensive sodium boiling tests have been carried out in two simulated LMFBR fuel pin bundles in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility at Oak Ridge National Laboratory. Experimental results from a 19-pin bundle (THORS Bundle 6A) have been previously reported, and experimental results from a 61-pin bundle (THORS Bundle 9) will be reported soon. The results discussed here are from the 19-pin bundle. Preliminary analysis has shown that the computational methods used and conclusions reached are equally valid for the 61-pin bundle, as well as the 19-pin in-reactor Sodium Loop Safety Facility (SLSF) W-1 experiment. The main result of THORS sodium boiling experimentation is that boiling behavior is determined by two-dimensional effects, i.e., the rates of mass, momentum and energy transfer in the direction perpendicular to the axes of the fuel pins.

  4. Thermal evaluation facility for LMFBR spent fuel transport

    SciTech Connect

    Wesley, D.A.

    1980-04-01

    A full-scale mock-up of a 217 pin breeder reactor fuel assembly in a cylindrical pipe was initially designed and constructed by Oak Ridge National Laboratory (ORNL). It was transferred to Sandia where it was extensively redesigned and modified. The 217 pin hexagonal core assembly was installed in a smaller diameter stainless steel pipe which more closely represents the diameter of a shipping canister or shipping cask basket wall. Two-hundred four of the tubes are electrically heated over an active length of 4-feet and the remaining thirteen are instrumented with multiple junction thermocouples which can be traversed axially. Thermocouples and heat-flux gauges are located on the hex core and canister perimeters at several axial locations.

  5. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    SciTech Connect

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

  6. LMFBR source term experiments in the Fuel Aerosol Simulant Test (FAST) facility

    SciTech Connect

    Petrykowski, J.C.; Longest, A.W.

    1985-01-01

    The transport of uranium dioxide (UO/sub 2/) aerosol through liquid sodium was studied in a series of ten experiments in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory (ORNL). The experiments were designed to provide a mechanistic basis for evaluating the radiological source term associated with a postulated, energetic core disruptive accident (CDA) in a liquid metal fast breeder reactor (LMFBR). Aerosol was generated by capacitor discharge vaporization of UO/sub 2/ pellets which were submerged in a sodium pool under an argon cover gas. Measurements of the pool and cover gas pressures were used to study the transport of aerosol contained by vapor bubbles within the pool. Samples of cover gas were filtered to determine the quantity of aerosol released from the pool. The depth at which the aerosol was generated was found to be the most critical parameter affecting release. The largest release was observed in the baseline experiment where the sample was vaporized above the sodium pool. In the nine ''undersodium'' experiments aerosol was generated beneath the surface of the pool at depths varying from 30 to 1060 mm. The mass of aerosol released from the pool was found to be a very small fraction of the original specimen. It appears that the bulk of aerosol was contained by bubbles which collapsed within the pool. 18 refs., 11 figs., 4 tabs.

  7. Steady natural convection heat transfer experiments in a horizontal annulus for the United States Spent Fuel Shipping Cask Technology Program. [LMFBR

    SciTech Connect

    Boyd, R. D.

    1981-04-01

    This experimental study deals with the measurement of the heat transfer across a horizontal annulus which is formed by an inner hexagonal cylinder and an outer concentric circular cylinder. The geometry simulates, in two dimensions, a liquid metal fast breeder reactor radioactive fuel subassembly inside a shipping container. This geometry is also similar to a radioactive fuel pin inside a horizontal reactor subassembly. The objective of the experiments is to measure the local and mean heat transfer at the surface of the inner hexagonal cylinder.

  8. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report

    SciTech Connect

    Todreas, N.E.; Golay, M.W.; Wold, L.

    1981-02-01

    Four tasks are reported on: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles. (DLC)

  9. Computation of the thermohydraulics in subassemblies for accident situations

    NASA Astrophysics Data System (ADS)

    Basque, G.

    Single-phase and two phase flow models of sodium boiling in an LMFBR are described. Results for single-phase calculations with mixed and natural convection, and for voiding of a subassembly are presented. A 2-D two-phase version of the computer code BACCHUS based on a homogeneous flow model with disequilibrium was tested. It yields results which are computationally stable and physically consistent. The validity of the porous body model for liquid flow is established by calculating representative single flows.

  10. Failure analysis of carbide fuels under transient overpower (TOP) conditions. [LMFBR

    SciTech Connect

    Nguyen, D.H.

    1980-06-01

    The failure of carbide fuels in the Fast Test Reactor (FTR) under Transient Overpower (TOP) conditions has been examined. The Beginning-of-Cycle Four (BOC-4) all-oxide base case, at $.50/sec ramp rate was selected as the reference case. A coupling between the advanced fuel performance code UNCLE-T and HCDA Code MELT-IIIA was necessary for the analysis. UNCLE-T was used to determine cladding failure and fuel preconditioning which served as initial conditions for MELT-III calculations. MELT-IIIA determined the time of molten fuel ejection from fuel pin.

  11. TEMP: a computer code to calculate fuel pin temperatures during a transient. [LMFBR

    SciTech Connect

    Bard, F E; Christensen, B Y; Gneiting, B C

    1980-04-01

    The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method.

  12. Preparation of carbide-type, advanced LMFBR fuel pellets for irradiation testing

    SciTech Connect

    Gutierrez, R.L.; Herbst, R.J.

    1980-06-01

    A carbothermic reduction process was established to fabricate single- and two-phase uranium-plutonium carbide fuel on a production basis. Sintering temperatures of 1550 and 1800/sup 0/C were used to prepare fuel densities of 98, 87, and 81% of theoretical.

  13. Technique for examining the fuel/cladding interface by TEM. [LMFBR

    SciTech Connect

    Yang, W.J.S.; Makenas, B.J.; Thomas, L.E.

    1983-05-01

    Fuel and fission-product interactions with the fuel-pin cladding is an area of concern and has been evaluated in the past principally by in-cell optical metallographic and electron-microprobe examinations. The applicability of three techniques for preparing specimens to reveal the microstructural details and local microchemistry of the fuel/cladding interface under conditions of high-resolution-scanning transmission-electron microscopy has been investigated. The specimen preparation techniques were designed to preserve the fuel/cladding interface and provide and maintain a specimen surface free from smearable alpha contamination. One of the techniques, Ni plating of a fuel cladding sample, preserved the entire cladding cross-section for examination. An Fe-oxide layer on the cladding inner surface was found in specimens prepared by this method. All three techniques of specimen preparation are described in some detail, along with their advantages and disadvantages.

  14. TWIST: a transient two-dimensional intra-subassembly thermal hydraulics model for LMFBRs

    SciTech Connect

    Khatib-Rahbar, M.; Cazzoli, E.G.

    1984-06-03

    Mathematical models and numerical methods for a two-dimensional porous body simulation of steady state and transient thermal-hydraulics conditions in LMFBR subassemblies resulting in the TWIST computer code are presented. Comparison of calculated results to steady state and transient out-of-pile sodium experiments show good agreement for cross-assembly temperature distributions for a wide range of heat transfer and flow conditions.

  15. Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system. [LMFBR

    SciTech Connect

    Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

    1980-05-01

    The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin.

  16. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1980-May 31, 1980

    SciTech Connect

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1980-01-01

    Experimental and theoretical work is reported on four tasks: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical local temperature files in LMFBR fuel rod bundles. (DLC)

  17. W-1 Sodium Loop Safety Facility experiment centerline fuel thermocouple performance. [LMFBR

    SciTech Connect

    Meyers, S.C.; Henderson, J.M.

    1980-05-01

    The W-1 Sodium Loop Safety Facility (SLSF) experiment is the fifth in a series of experiments sponsored by the Department of Energy (DOE) as part of the National Fast Breeder Reactor (FBR) Safety Assurance Program. The experiments are being conducted under the direction of Argonne National Laboratory (ANL) and Hanford Engineering Development Laboratory (HEDL). The irradiation phase of the W-1 SLSF experiment was conducted between May 27 and July 20, 1979, and terminated with incipient fuel pin cladding failure during the final boiling transient. Experimental hardware and facility performed as designed, allowing completion of all planned tests and test objectives. This paper focuses on high temperature in-fuel thermocouples and discusses their development, fabrication, and performance in the W-1 experiment.

  18. Two dimensional, two fluid model for sodium boiling in LMFBR fuel assemblies

    SciTech Connect

    Granziera, M.R.; Kazimi, M.S.

    1980-05-01

    A two dimensional numerical model for the simulation of sodium boiling transient was developed using the two fluid set of conservation equations. A semiimplicit numerical differencing scheme capable of handling the problems associated with the ill-posedness implied by the complex characteristic roots of the two fluid problems was used, which took advantage of the dumping effect of the exchange terms. Of particular interest in the development of the model was the identification of the numerical problems caused by the strong disparity between the axial and radial dimensions of fuel assemblies. A solution to this problem was found which uses the particular geometry of fuel assemblies to accelerate the convergence of the iterative technique used in the model. Three sodium boiling experiments were simulated with the model, with good agreement between the experimental results and the model predictions.

  19. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    SciTech Connect

    R. Wigeland; K. Hamman

    2009-09-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving

  20. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    SciTech Connect

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

  1. Investigation of the physical and numerical foundations of two-fluid representation of sodium boiling with applications to LMFBR experiments

    SciTech Connect

    No, H.C.; Kazimi, M.S.

    1983-03-01

    This work involves the development of physical models for the constitutive relations of a two-fluid, three-dimensional sodium boiling code, THERMIT-6S. The code is equipped with a fluid conduction model, a fuel pin model, and a subassembly wall model suitable for stimulating LMFBR transient events. Mathematically rigorous derivations of time-volume averaged conservation equations are used to establish the differential equations of THERMIT-6S. These equations are then discretized in a manner identical to the original THERMIT code. A virtual mass term is incorporated in THERMIT-6S to solve the ill-posed problem. Based on a simplified flow regime, namely cocurrent annular flow, constitutive relations for two-phase flow of sodium are derived. The wall heat transfer coefficient is based on momentum-heat transfer analogy and a logarithmic law for liquid film velocity distribution. A broad literature review is given for two-phase friction factors. It is concluded that entrainment can account for some of the discrepancies in the literature. Mass and energy exchanges are modelled by generalization of the turbulent flux concept. Interfacial drag coefficients are derived for annular flows with entrainment. Code assessment is performed by simulating three experiments for low flow-high power accidents and one experiment for low flow/low power accidents in the LMFBR. While the numerical results for pre-dryout are in good agreement with the data, those for post-dryout reveal the need for improvement of the physical models. The benefits of two-dimensional non-equilibrium representation of sodium boiling are studied.

  2. General formulation of an HCDA bubble rising in a sodium pool and the effect of nonequilibrium on fuel transport. [LMFBR

    SciTech Connect

    Kocamustafaogullari, G.; Chan, S.H.

    1980-01-01

    This report which improved the formulation of the previous reports is designed to investigate the effect of the interfacial nonequilibrium mass transfer and the radiative heat transfer on the amount of the fuel vapor condensed before the bubble reaches to the cover-gas region. Consideration is given to a fuel dominated bubble which is assumed to have just penetrated into the sodium pool in a spherical form subsequent to an Hypothetical Core Disruptive Accident (HCDA). The two-phase bubble mixture as it rises through the sodium pool to the cover-gas region is formulated. The formulation takes into account the effects of the nonequilibrium mass transfer at the interfaces and of the radiative cooling of the bubble as well as the kinematic, dynamic and thermal effects of the surrounding fields. The results of calculation for the amount of the fuel vapor condensed before the bubble reaches the cover-gas region are presented over a wide possible range of the evaporation coefficient as well as the liquid sodium-bubble interface absorbtivity.

  3. Assembly planning based on subassembly extraction

    NASA Technical Reports Server (NTRS)

    Lee, Sukhan; Shin, Yeong Gil

    1990-01-01

    A method is presented for the automatic determination of assembly partial orders from a liaison graph representation of an assembly through the extraction of preferred subassemblies. In particular, the authors show how to select a set of tentative subassemblies by decomposing a liaison graph into a set of subgraphs based on feasibility and difficulty of disassembly, how to evaluate each of the tentative subassemblies in terms of assembly cost using the subassembly selection indices, and how to construct a hierarchical partial order graph (HPOG) as an assembly plan. The method provides an approach to assembly planning by identifying spatial parallelism in assembly as a means of constructing temporal relationships among assembly operations and solves the problem of finding a cost-effective assembly plan in a flexible environment. A case study of the assembly planning of a mechanical assembly is presented.

  4. Model for LMFBR core transient analysis in real-time

    SciTech Connect

    Tzanos, C.P.

    1986-01-01

    This paper discusses the modeling of LMFBR core transients. It is shown that with a proper choice of shape functions a nodal approximation of the coolant, cladding, and fuel temperature distributions leads to adequately accurate power and temperature predictions, as well as adequately short computation times.

  5. Transient Response in LMFBR System.

    Energy Science and Technology Software Center (ESTSC)

    1999-04-26

    SSC-L (the Super System Code) calculates the thermohydraulic response of loop-type liquid metal fast breeder reactor (LMFBR) systems during operational, incidental, and accidental transients, especially natural circulation events. Modules simulated and parameters calculated include: core flow rates and temperatures, loop flow rates and temperatures, pump performance, and heat exchanger operation. Additionally, SSC-L accounts for all plant protection and plant control systems. Although the primary emphasis is on transients for safety analysis, SSC-L can be usedmore » for many other applications, such as scoping analysis for plant design and specification of various components. Any number of user-specified loops, pipes, and nodes are permitted. Both single- and two-phase thermal-hydraulics are used in a multi-channel core representation. Inter-assembly flow redistribution is accounted for using a detailed fuel pin model. The heat transport system geometry is user-specified. SSC-L provides steady-state and transient options and a restart capability. Input is free format in a modular structure that makes use of abstract data management techniques.« less

  6. Ejector subassembly for dual wall air drilling

    SciTech Connect

    Kolle, J.J.

    1996-09-01

    The dry drilling system developed for the Yucca Mountain Site Characterization Project incorporates a surface vacuum system to prevent drilling air and cuttings from contaminating the borehole wall during coring operations. As the drilling depth increases, however there is a potential for borehole contamination because of the limited volume of air which can be removed by the vacuum system. A feasibility analysis has shown that an ejector subassembly mounted in the drill string above the core barrel could significantly enhance the depth capacity of the dry drilling system. The ejector subassembly would use a portion of the air supplied to the core bit to maintain a vacuum on the hole bottom. The results of a design study including performance testing of laboratory scale ejector simulator are presented here.

  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. LMFBR models for the ORIGEN2 computer code

    SciTech Connect

    Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

    1981-10-01

    Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th-/sup 238/U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given.

  9. LMFBR models for the ORIGEN2 computer code

    SciTech Connect

    Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

    1983-06-01

    Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th-/sup 233/U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given.

  10. Technical feasibility of transition phase tests in TREAT. [LMFBR

    SciTech Connect

    Stewart, R.R.; Bauer, T.H.; Hoff, O.I.; Koyama, K.; Stephenson, M.E.; Kraft, T.E.

    1982-01-01

    Understanding the behavior of molten fuel-steel mixtures subjected to fission heating within a HCDA environment is essential to continuing the mechanistic description of the whole-core accident into the transition phase, and further to a permanent subcritical and safe fuel debris configuration. Fundamentally, the RX1 TREAT test will simulte the transition phase of a HCDA (the accident phase in which the fuel in individual subassemblies melts and becomes a heat-generating pool of molten fuel and boiling steel). This assessment of the feasibility of such a test indicates that a transition phase test can be achieved in TREAT, at power levels simulating decay heat.

  11. Isolation of Bacterial Type IV Machine Subassemblies

    PubMed Central

    Sarkar, Mayukh K.; Husnain, Seyyed I.; Jakubowski, Simon J.; Christie, Peter J.

    2012-01-01

    The bacterial type IV secretion systems (T4SSs) deliver DNA and protein substrates to bacterial and eukaryotic target cells generally by a mechanism requiring direct contact between donor and target cells. Recent advances in defining the architectures of T4SSs have been made through isolation of machine sub-assemblies for further biochemical and ultrastructural analysis. Here, we describe a protocol for isolation and characterization of VirB protein complexes from the paradigmatic VirB/VirD4 T4SS of Agrobacterium tumefaciens. This protocol can be adapted for isolation of T4SS subassemblies from other gram-negative bacteria as well as gram-positive bacteria. The biological importance of isolated T4SS subcomplexes can be assessed by assaying for copurification of trapped or cross-linked substrates. This can be achieved with a modified form of the chromatin immunoprecipitation (ChIP) assay termed transfer DNA immunoprecipitation (TrIP). Here, a TrIP protocol is described for recovery of formaldehyde-cross-linked DNA substrate–channel subunit complexes from cells employing T4SSs for conjugative DNA transfer. PMID:23299736

  12. Structural response of reactor-core hexcan subassemblies subjected to dynamic overpressurization under accident conditions

    SciTech Connect

    Pfeiffer, P.A.; Kulak, R.F.

    1993-01-01

    This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall.

  13. Structural response of reactor-core hexcan subassemblies subjected to dynamic overpressurization under accident conditions

    SciTech Connect

    Pfeiffer, P.A.; Kulak, R.F.

    1993-06-01

    This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall.

  14. Cost-reduction potential in LMFBR design

    SciTech Connect

    Chang, Y.I.; Till, C.E.

    1983-01-01

    LWR capital costs have escalated continuously over the years to the point where today its economics represent a bar to further LWR deployment in the US. High initial costs and the promise of a similar pattern of cost escalation in succeeding years for the LMFBR would effectively stop LMFBR deployment in this country before it could even begin. LWR cost escalation in the main can be traced to large increases in both amounts and unit costs of construction materials and to greatly lengthened construction times. Innovative approaches to LMFBR design are now being pursued that show promise for substantial cost reductions particularly in those areas that have contributed most to LWR cost increases.

  15. 3. INTERIOR VIEW, LOOKING EAST, WITH SUBASSEMBLY OF END SECTION ...

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

    3. INTERIOR VIEW, LOOKING EAST, WITH SUBASSEMBLY OF END SECTION SHEAR PLATE FOR HOPPER CAR AND NINA CASTLE, CLASS A WELDER. - Pullman Standard Company Plant, Fabrication Assembly Shop, 401 North Twenty-fourth Street, Bessemer, Jefferson County, AL

  16. 4. INTERIOR VIEW, LOOKING EAST, WITH SUBASSEMBLY OF END SECTION ...

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

    4. INTERIOR VIEW, LOOKING EAST, WITH SUBASSEMBLY OF END SECTION SHEAR PLATE FOR HOPPER CAR AND NINA CASTLE, CLASS A WELDER. - Pullman Standard Company Plant, Fabrication Assembly Shop, 401 North Twenty-fourth Street, Bessemer, Jefferson County, AL

  17. Design of a singularity-free articulated arm subassembly

    SciTech Connect

    Remis, S.J.; Stanisic, M.M. . Aerospace and Mechanical Engineering Dept.)

    1993-12-01

    Adding a redundant degree of freedom to the shoulder pointing system complex of an articulated arm subassembly makes it possible to achieve a maximal workspace that is free of singularities. This paper derives a functional constraint between three of the four joints of this new type of arm, achieving a singularity-free workspace encompassing the entire reachable volume between the maximal- and minimal-reach surfaces. The large volume of dexterous workspace is verified by animation of the resulting arm design. Graphical results from the animation are presented comparing the dexterous workspace of this new arm to that of the standard nonredundant articulated arm subassembly such as found in the Puma manipulator.

  18. STRUCTURE FOR SUB-ASSEMBLIES OF ELECTRONIC EQUIPMENT

    DOEpatents

    Bell, P.R.; Harris, C.C.

    1959-03-31

    Sub-assemblies for electronic systems, particularly a unit which is self- contained and which may be adapted for quick application to and detachment from a chassis or panel, are discussed. The disclosed structure serves the dual purpose of a cover or enclosure for a subassembly comprising a base plate and also acts as a clamp for retaining the base plate in position on a chassis. The clamping action is provided by flexible fingers projecting from the side walls of the cover and extending through grooves in the base plate to engage with the opposite side of the chassis.

  19. Force-Based Reasoning for Assembly Planning and Subassembly Stability Analysis

    NASA Technical Reports Server (NTRS)

    Lee, S.; Yi, C.; Wang, F-C.

    1993-01-01

    In this paper, we show that force-based reasoning, for identifying a cluster of parts that can be decomposed naturally by the applied force, plays an important role in selecting feasible subassemblies and analyzing subassembly stability in assembly planning.

  20. 2. INTERIOR VIEW, LOOKING EAST, OF TRACK TWO WITH SUBASSEMBLY ...

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

    2. INTERIOR VIEW, LOOKING EAST, OF TRACK TWO WITH SUBASSEMBLY OF END SECTION SHEAR PLATE FOR ROUND-SIDE HOPPER CAR (DESIGNED FOR TRANSPORT OF PLASTIC PELLETS FOR PETROCHEMICAL INDUSTRY) AND DEXTER WALTON, CLASS A WELDER. - Pullman Standard Company Plant, Fabrication Assembly Shop, 401 North Twenty-fourth Street, Bessemer, Jefferson County, AL

  1. Aberration analysis and efficiency improvement of a bidirectional optical subassembly

    NASA Astrophysics Data System (ADS)

    Wu, Hao; Huang, Zhangdi; Yu, Ziyan; Qian, Xiaoshi; Xu, Fei; Chen, Beckham; Lu, Yanqing

    2009-10-01

    An approach to improve the coupling efficiency of bidirectional optical subassembly (BOSA) modules is proposed and experimentally demonstrated. We analyzed the wavefront aberration coefficients of a typical BOSA. It was found that the 45-deg wavelength filter induces coma and astigmatism, and then it further deteriorates the laser diode to fiber coupling. We measured the BOSA efficiencies based on a series of different filters. For a typical 0.5-mm filter, 25% coupling efficiency improvement was achieved by optimizing the filter parameters.

  2. Static regenerative fuel cell system for use in space

    NASA Technical Reports Server (NTRS)

    Levy, Alexander H. (Inventor); VanDine, Leslie L. (Inventor); Trocciola, John C. (Inventor)

    1989-01-01

    The cell stack can be operated as a fuel cell stack or as an electrolysis cell stack. The stack consists of a series of alternate fuel cell subassemblies with intervening electrolysis cell subassemblies, and interspersed cooling plates. The water produced and consumed in the two modes of operation migrates between adjacent cell subassemblies. The component plates are annular with a central hydrogen plenum and integral internal oxygen manifolds. No fluid pumps are needed to operate the stack in either mode.

  3. LMFBR with booster pump in pumping loop

    DOEpatents

    Rubinstein, H.J.

    1975-10-14

    A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation.

  4. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, W.E.; Trapp, T.J.

    1983-06-10

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  5. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, William E.; Trapp, Turner J.

    1985-01-01

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  6. Vibration test plan for a space station heat pipe subassembly

    SciTech Connect

    Parekh, M.B.

    1987-09-29

    This test plan describes the Sundstrand portion of task two of Los Alamos National Laboratory (LANL) contract 9-x6H-8102L-1. Sundstrand Energy Systems was awarded a contract to investigate the performance capabilities of a potassium liquid metal heat pipe as applied to the Organic Rankine Cycle (ORC) solar dynamic power system for the Space Station. The test objective is to expose the heat pipe subassembly to the random vibration environment which simulates the space shuttle launch condition. The results of the test will then be used to modify as required future designs of the heat pipe.

  7. 30 CFR 27.35 - Tests to determine life of critical components and subassemblies.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Tests to determine life of critical components and subassemblies. 27.35 Section 27.35 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION... Requirements § 27.35 Tests to determine life of critical components and subassemblies. Replaceable...

  8. Building and testing of MIDAS instrument sub-assemblies

    NASA Astrophysics Data System (ADS)

    Lewis, S. D.

    2001-09-01

    The MIDAS instrument is an atomic force microscope developed by ESTEC to fly on Rosetta. The purpose of the instrument is to sample and characterise cometary dust, which impinges upon a facetted wheel contained within the instrument enclosure. Due to its relative complexity, the long cruise phase of the Rosetta mission and the relatively novel use of piezomotors for all drive requirements the instrument has a number of interesting mechanisms engineering challenges. This paper describes the lubricant selection, EM and FM subassembly build and test campaigns carried out by AEA Technology Space in close support of the instrumentlevel activities which ran in parallel at ESTEC. The paper also identifies some lessons learned, which can be generally applied in other mechanism programmes.

  9. Generation IV nuclear energy system initiative. Pin core subassembly designfor the Gas-Cooled Fast Reactor.

    SciTech Connect

    Farmer, M. T.; Hoffman, E. A.; Pfeiffer, P. F.; Therios, I. U.

    2006-07-31

    The Gas-Cooled Fast Reactor (GFR) is one of six systems selected for viability assessment in the Generation IV program. It features a closed nuclear fuel cycle, consisting of a high-temperature helium-cooled fast spectrum reactor, coupled to a direct-cycle helium turbine for electricity production. The GFR combines the advances of fast spectrum systems with those of high-temperature systems. It was clear from the very beginning that GFR design should be driven by the objective to offer a complementary approach to liquid metal cooling. On this basis, CEA and the US DOE decided to collaborate on the pre-conceptual design of a GFR. This reactor design will provide a high level of safety and full recycling of the actinides, and will also be highly proliferation resistant and economically attractive. The status of this collaborative project is that two unit sizes, 600 MWt and 2400 MWt were selected as the focus of the design and safety studies. Researchers studied fuel forms, fuel assembly/element designs, core configurations, primary and balance-of-plant layouts, and safety approaches for both of these unit sizes. Results regarding the feasibility of this GFR design are encouraging. For example, sustainability and non-proliferation goals can be met and the proposed concept has attractive safety features. These features take advantage of the helium in terms of its neutronic quasi-transparency as well as the enhanced Doppler effect in connection with candidate fuel and structural materials. The current design trend is to consider high unit power for the GFR (2400 MWt), an attractive level for the power density (100 MW/m{sup 3}), and the implementation of an innovative plate type fuel or pin type sub-assembly with carbide-based actinide compounds and SiC-based structural materials. Work is still needed to refine the safety approach, to select the main system options, and to more definitively establish economic parameters.

  10. A study on reactor core failure thresholds to safety operation of LMFBR

    SciTech Connect

    Kazuo, Haga; Hiroshi, Endo; Tomoko, Ishizu; Yoshihisa, Shindo

    2006-07-01

    Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

  11. Reactor control rod timing system. [LMFBR

    DOEpatents

    Wu, P.T.K.

    1980-03-18

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  12. SACO-1: a fast-running LMFBR accident-analysis code

    SciTech Connect

    Mueller, C.J.; Cahalan, J.E.; Vaurio, J.K.

    1980-01-01

    SACO is a fast-running computer code that simulates hypothetical accidents in liquid-metal fast breeder reactors to the point of permanent subcriticality or to the initiation of a prompt-critical excursion. In the tradition of the SAS codes, each subassembly is modeled by a representative fuel pin with three distinct axial regions to simulate the blanket and core regions. However, analytic and integral models are used wherever possible to cut down the computing time and storage requirements. The physical models and basic equations are described in detail. Comparisons of SACO results to analogous SAS3D results comprise the qualifications of SACO and are illustrated and discussed.

  13. Development of a methodology for analysis of delayed-neutron signals. [LMFBR

    SciTech Connect

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

    1980-02-01

    Experimental and analytical techniques have been developed for analysis and characterization of delayed-neutron (DN) signals that can provide diagnostic information to augment data from cover-gas analyses in the detection and identification of breached elements in an LMFBR. Eleven flow-reduction tests have been run in EBR-II to provide base data support for predicting DN signal characteristics during exposed-fuel operation. Results from the tests demonstrate the feasibility and practicability of response-analysis techniques for determining (a) the transit time, T/sub tr/, for DN emitters traveling from the core to the detector and (b) the isotropic holdup time, T/sub h/, of DN precursors in the fuel element.

  14. GE post-test analysis of SLSF experiment W-1 through LOPI-4. [LMFBR

    SciTech Connect

    Gregoire, K.E.; Atcheson, D.B.; Knight, D.D.

    1981-07-01

    SLSF experiment W-1 was designed to investigate fuel pin-to-coolant heat transfer during various LMFBR flow-coastdown events in the burnup interval from 0.0 atom percent to 0.5 atom percent. In the study reported here, data from in-fuel thermocouples and coolant (wire-wrap) thermocouples were evaluated during steady-state and transient operation from the beginning of the experiment through LOPI-4 (Loss-of-Piping-Integrity Transient Number 4). The objective of the data evaluation was to determine how maximum coolant temperatures during successive LOPI transients were affected by burnup. A second objective was to identify the mechanisms responsible for this burnup effect.

  15. Comments on US LMFBR steam generator base technology

    SciTech Connect

    Simmons, W.R.

    1984-01-01

    The development of steam generators for the LMFBR was recognized from the onset by the AEC, now DOE, as a difficult, challenging, and high-priority task. The highly reactive nature of sodium with water/steam requires that the sodium-water/steam boundaries of LMFBR steam generators possess a degree of leak-tightness reliability not normally attempted on a commercial scale. In addition, the LMFBR steam generator is subjected to high fluid temperatures and severe thermal transients. These requirements place great demand on materials, fabrication processes, and inspection methods; and even greater demands on the designer to provide steam generators that can meet these demanding requirements, be fabricated without unreasonable shop requirements, and tolerate off-normal effects.

  16. Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

    SciTech Connect

    Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

    1982-01-01

    The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling.

  17. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

    SciTech Connect

    Todreas, N.E.; Cheng, S.K.; Basehore, K.

    1984-08-01

    This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.

  18. Fluid-mixing studies in a hexagonal 217-pin wire-wrapped rod bundle. [LMFBR

    SciTech Connect

    Symolon, P.D.; Todreas, N.E.

    1981-02-01

    Mixing, pressure drop, and flow split experiments were performed on a 217 pin LMFBR fuel bundle with a pitch to diameter ratio of 1.25 and a lead length of 12 inches. It was found that the turbulent flow data could best be characterized by the energy parameter C/sub 1L/=.106, which is 9% higher than the value from the correlation of Chiu et al. Chiu's correlation was developed on a data base of 61 and 91 pins. The spread of existing data about the correlation is +- 25%, but the error band on our data is expected to be less (approx. +- 10% since injection depth effects were not previously considered). This result is consistent with the concept of increased swirl flow in larger bundles (more pins).

  19. Study on mixed convective flow penetration into subassembly from reactor hot plenum in FBRs

    SciTech Connect

    Kobayashi, J.; Ohshima, H.; Kamide, H.; Ieda, Y.

    1995-09-01

    Fundamental experiments using water were carried out in order to reveal the phenomenon of mixed convective flow penetration into subassemblies from a reactor`s upper plenum of fast breeder reactors. This phenomenon appears under a certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system for decay heat removal and might influence the natural circulation head which determines the core flow rate and therefore affects the core coolability. In the experiment, a simplified model which simulates an upper plenum and a subassembly was used and the ultrasonic velocity profile monitor as well as thermocouples were applied for the simultaneous measurement of velocity and temperature distributions in the subassembly. From the measured data, empirical equations related to the penetration flow onset condition and the penetration depth were obtained using relevant parameters which were derived from dimensional analysis.

  20. Application of Autoregressive Models to In-Service Estimation of Transient Response for LMFBR Process Instrumentation

    SciTech Connect

    Ueda, Masashi; Tomobe, Katsuma; Setoguchi, Keiichi; Endou, Akira

    2002-02-15

    The response of a sensor depends on its operating conditions, and thus it is desirable to develop an in-service method for response time estimation. The applicability of the autoregressive (AR) model for this purpose was examined in the case of the fuel subassembly outlet coolant thermocouples and the primary circuit electromagnetic flowmeter (EMF) of Monju, the prototype fast breeder reactor in Japan.The use of an AR model with exogenous input (ARX model) is possible when the physical variable to be sensed can be observed by an alternative means with a faster response time than that of the sensor in question. In the case of the subassembly outlet thermocouple, the temperature output from an eddy-current sensor, during pseudorandom reactor power variation, served as the exogenous input.In respect to the thermocouple response, AR and ARX modeling were shown to be applicable, and the transient responses thus derived agreed well with each other and with the results measured by means of a step change in sodium temperature. However, the primary circuit EMF response time, estimated using the AR model, decreased with increasing flow rate even when approaching the rated flow, demonstrating that the method was not completely applicable. Nevertheless, it can be concluded that the response is faster than that estimated in the rated condition.

  1. COBRA-PI: an extension of the COBRA-3M code dynamically dimensioned to accept pin bundles of any size. [LMFBR

    SciTech Connect

    Froehle, P.H.; Bauer, T.H.

    1983-03-28

    COBRA, in general, performs a thermal-hydraulic analysis of an actual pin bundle by subdividing the bundle cross-section into coolant subchannels, pin sectors, duct wall sectors. Its calculation includes heat convected axially upward through coolant mass flow, heat flow between pin sectors and adjoining subchannels, and heat and mass flow between coolant subchannels. COBRA-3M is a version of COBRA built for LMFBR applications, that includes a sophisticated thermal model of fuel pins and duct wall. COBRA-3M that can explicitly model a wider variety of pin bundle configurations than 3M would allow and includes significant improvements to its thermal modeling. COBRA-PI is currently being used for thermal-hydraulic analysis of hypothetical LMFBR accident transients in both power and flow. Pin bundles currently being analyzed explicitly range from 7 to 37 pins of axial lengths ranging from approx. 0.3-2.0 meters.

  2. Hot Fuel Examination Facility's neutron radiography reactor

    SciTech Connect

    Pruett, D.P.; Richards, W.J.; Heidel, C.C.

    1983-01-01

    Argonne National Laboratory-West is located near Idaho Falls, Idaho, and is operated by the University of Chicago for the United States Department of Energy in support of the Liquid Metal Fast Breeder Reactor Program, LMFBR. The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both nondestructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the nondestructive examination techniques utilized at HFEF is neutron radiography, which is provided by the NRAD reactor facility (a TRIGA type reactor) below the HFEF hot cell.

  3. Indirect detection of dryout in simulated LMFBR fuel assemblies

    SciTech Connect

    Levin, A.E.

    1981-01-01

    The method of indirect dryout detection was developed by an examination of the data from THORS Bundle 6A. This was a 19-pin bundle of FFTF configuration. The pin size, wire-wrap size and axial pitch were identical to those in Bundle 9; the major difference between the Bundle 6A and Bundle 9 FPSs was the length of the upper unheated zone, which simulated, in Bundle 6A, the reflector and fission gas plenum in FFTF (1.19 m) and, in Bundle 9, the upper axial blanket and fission gas plenum in CRBR (1.54 m). In addition, Bundle 6A had half-size (0.71 mm) edge channel wire-wraps and a low thermal inertia (0.51 mm thick) duct wall surrounded by calcium silicate insulation in an attempt to flatten the bundle temperature profile.

  4. Hydrodynamics of large-scale fuel-coolant interactions. [LMFBR

    SciTech Connect

    Baines, M.; Board, S.J.; Buttery, N.E.

    1980-06-01

    The analogy between thermal reactive and chemical reactive flows suggests that all propagating thermal explosions have a detonation-like (i.e., shock) structure. A vapor detonation model, which allows for thermal disequilibrium in the coolant, is developed. It is suggested that similar nonequilibrium effects may limit the efficiency of UO/sub 2/-sodium system, however, because of high conductivity of the coolant. 34 refs.

  5. International Space Station Alpha trace contaminant control subassembly life test report

    NASA Technical Reports Server (NTRS)

    Tatara, J. D.; Perry, J. L.

    1995-01-01

    The Environmental Control and Life Support System (ECLSS) Life Test Program (ELTP) began with Trace Contaminant Control Subassembly (TCCS) Life Testing on November 9, 1992, at 0745. The purpose of the test, as stated in the NASA document 'Requirements for Trace Contaminant Control Subassembly High Temperature Catalytic Oxidizer Life Testing (Revision A)' was to 'provide for the long duration operation of the ECLSS TCCS HTCO (High Temperature Catalytic Oxidizer) at normal operating conditions... (and thus)... to determine the useful life of ECLSS hardware for use on long duration manned space missions.' Specifically, the test was designed to demonstrate thermal stability of the HTCO catalyst. The report details TCCS stability throughout the test. Graphs are included to aid in evaluating trends and subsystem anomalies. The report summarizes activities through the final day of testing, January 17, 1995 (test day 762).

  6. Valve for fuel pin loading system

    DOEpatents

    Christiansen, David W.

    1985-01-01

    A cyclone valve surrounds a wall opening through which cladding is projected. An axial valve inlet surrounds the cladding. Air is drawn through the inlet by a cyclone stream within the valve. An inflatable seal is included to physically engage a fuel pin subassembly during loading of fuel pellets.

  7. Valve for fuel pin loading system

    DOEpatents

    Christiansen, D.W.

    1984-01-01

    A cyclone valve surrounds a wall opening through which cladding is projected. An axial valve inlet surrounds the cladding. Air is drawn through the inlet by a cyclone stream within the valve. An inflatable seal is included to physically engage a fuel pin subassembly during loading of fuel pellets.

  8. Dynamic stability experiments in sodium-heated steam generators. [LMFBR

    SciTech Connect

    France, D.M.; Roy, R.; Carlson, R.D.; Chiang, T.

    1984-01-01

    Seventy-two dynamic stability tests were performed in the sodium-heated boiling-water test facility at Argonne National Laboratory. A full-scale LMFBR steam generator tube was employed as the test section operating over the water parameter ranges of 6.9 to 15.9 MPa pressure and 170 to 800 kg/m/sup 2/.s mass flux. The stability thresholds from the test compared well to the predictions of a modified version of a correlation equation recently published by other investigators. Typical experimental data and the modified correlation equation are presented.

  9. Pressurized solid oxide fuel cell integral air accumular containment

    DOEpatents

    Gillett, James E.; Zafred, Paolo R.; Basel, Richard A.

    2004-02-10

    A fuel cell generator apparatus contains at least one fuel cell subassembly module in a module housing, where the housing is surrounded by a pressure vessel such that there is an air accumulator space, where the apparatus is associated with an air compressor of a turbine/generator/air compressor system, where pressurized air from the compressor passes into the space and occupies the space and then flows to the fuel cells in the subassembly module, where the air accumulation space provides an accumulator to control any unreacted fuel gas that might flow from the module.

  10. New approach to the design of core support structures for large LMFBR plants

    SciTech Connect

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-01-01

    The paper describes an innovative design concept for a LMFBR Core Support Structure. A hanging Core Support Structure is described and analyzed. The design offers inherent safety features, constructibility advantages, and potential cost reductions.

  11. Thermal analysis of a six-channel heat-generating blockage in an LMFBR

    SciTech Connect

    Warinner, D.K.; Chao, D.H.Y.

    1980-01-01

    This paper presents a case study of the temperature fields within and around a six-channel blockage designed as a molten-fuel-release initiator in SLSF-P4, an in-reactor experiment (37-mixed-oxide pin bundle) planned for February, 1981, irradiation. To meet the experiment objectives, a minimum of ten grams of molten UO/sub 2/ must be ejected into the sodium stream from one, two, or three such blockages. The temperature fields of the electrodeposited-nickel blockage filled with a mixture of UO/sub 2/ powder, stainless steel, and gas are found at intervals of full power. The SS content, type of gas, and porosity were parameters varied in this study which used the computer codes THYME-B, SABRE-1, and ANL's version of THTB. State-of-the-art treatments of the conductivity of the mixture and the gas-gap conductance are included. The contrived-blockage design has been found to maintain structural integrity until sufficient molten fuel exists to release, challenge the subassembly, and be detected by delayed-neutron and fission-product monitors. This will serve to resolve lingering questions on rapid pin-to-pin propagation, blockage propagation, and other local-fault issues.

  12. Calculation of Doses Due to Accidentally Released Plutonium From An LMFBR

    SciTech Connect

    Fish, B.R.

    2001-08-07

    Experimental data and analytical models that should be considered in assessing the transport properties of plutonium aerosols following a hypothetical reactor accident have been examined. Behaviors of released airborne materials within the reactor containment systems, as well as in the atmosphere near the reactor site boundaries, have been semiquantitatively predicted from experimental data and analytical models. The fundamental chemistry of plutonium as it may be applied in biological systems has been used to prepare models related to the intake and metabolism of plutonium dioxide, the fuel material of interest. Attempts have been made to calculate the possible doses from plutonium aerosols for a typical analyzed release in order to evaluate the magnitude of the internal exposure hazards that might exist in the vicinity of the reactor after a hypothetical LMFBR (Liquid-Metal Fast Breeder Reactor) accident. Intake of plutonium (using data for {sup 239}Pu as an example) and its distribution in the body were treated parametrically without regard to the details of transport pathways in the environment. To the extent possible, dose-response data and models have been reviewed, and an assessment of their adequacy has been made so that recommended or preferred practices could be developed.

  13. Seventeen years of LMFBR experience: Experimental Breeder Reactor II (EBR-II)

    SciTech Connect

    Perry, W.H.; Lentz, G.L.; Richardson, W.J.; Wolz, G.C.

    1982-05-01

    Operating experience at EBR-II over the past 17 years has shown that a sodium-cooled pool-type reactor can be safely and efficiently operated and maintained. The reactor has performed predictably and benignly during normal operation and during both unplanned and planned plant upsets. The duplex-tube evaporators and superheaters have never experienced a sodium/water leak, and the rest of the steam-generating system has operated without incident. There has been no noticeable degradation of the heat transfer efficiency of the evaporators and superheaters, except for the one superheater replaced in 1981. There has been no need to perform any chemical cleaning of steam-system components. Operation of EBR-II has produced a wealth of information. As an irradiation facility, EBR-II has generated specific information on the behavior of oxide, carbide, and metallic fuels. As an LMFBR power plant, EBR-II has produced general information related to plant-systems and equipment design, plant safety, plant availability, and plant maintenance.

  14. Fatigue of LMFBR piping due to flow stratification

    SciTech Connect

    Woodward, W.S.

    1983-01-01

    Flow stratification due to reverse flow was simulated in a 1/5-scale water model of a LMFBR primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. The behavior of the interface was characterized in terms of location, local temperature fluctuation and duration for various reverse flow conditions. A structural assessment was performed to determine the effects of stratified flow on the fatigue life of the pipe. Both the static and dynamic aspects of flow stratification were examined. The dynamic interface produces thermal striping on the inside of the pipe wall which is shown to have the most deleterious effect on the pipe wall and produce significant fatigue damage relative to a static interface.

  15. MIT LMFBR blanket research project. Final summary report

    SciTech Connect

    Driscoll, M.J.

    1983-08-01

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record.

  16. Simple LMFBR axial-flow friction-factor correlation

    SciTech Connect

    Chan, Y.N.; Todreas, N.E.

    1982-12-01

    Complicated LMFBR axial lead-length averaged friction-factor correlations are reduced to an easy, ready-to-use function of bundle Reynolds number for wire-wrapped bundles. The function together with the power curves to calculate the associated constants are incorporated in a computer preprocessor, EZFRIC. The constants required for the calculation of the subchannels and bundle friction factors are derived and correlated into power curves of geometrical parameters. A computer program, FRIC, which can alternatively be used to accurately calculate these constants is also included. The accurate values of the constants and the corresponding values predicted by the power curves and percentage error of prediction are tabulated for a wide variety of geometries of interest.

  17. Breeding nuclear fuels with accelerators: replacement for breeder reactors

    SciTech Connect

    Grand, P.; Takahashi, H.

    1984-01-01

    One application of high energy particle accelerators has been, and still is, the production of nuclear fuel for the nuclear energy industry; tantalizing because it would create a whole new industry. This approach to producing fissile from fertile material was first considered in the early 1950's in the context of the nuclear weapons program. A considerable development effort was expended before discovery of uranium ore in New Mexico put an end to the project. Later, US commitment to the Liquid Metal Fast Breeder Reactors (LMFBR) killed any further interest in pursuing accelerator breeder technology. Interest in the application of accelerators to breed nuclear fuels, and possibly burn nuclear wastes, revived in the late 1970's, when the LMFBR came under attack during the Carter administration. This period gave the opportunity to revisit the concept in view of the present state of the technology. This evaluation and the extensive calculational modeling of target designs that have been carried out are promising. In fact, a nuclear fuel cycle of Light Water Reactors and Accelerator Breeders is competitive to that of the LMFBR. At this time, however, the relative abundance of uranium reserves vs electricity demand and projected growth rate render this study purely academic. It will be for the next generation of accelerator builders to demonstate the competitiveness of this technology versus that of other nuclear fuel cycles, such as LMFBR's or Fusion Hybrid systems. 22 references, 1 figure, 5 tables.

  18. DEVELOPMENT OF DESIGN TOOLS TO FACILITATE/PROMOTE SUSTAINABLE DESIGN OF PROTON EXCHANGE MEMBRANE FUEL CELLS

    EPA Science Inventory

    Objective is to develop and demonstrate 2 sets of of design tools that are applicable to the manufacture of proton exchange membrane fuel cell systems. First set will offer guidance to fuel cell designers for end of life options suited to subassembly. Second set will give fuel ...

  19. The detonation electric effect as applied to the MC-2453 driver subassembly. Progress report, October 1971--December 1971

    SciTech Connect

    Boettner, J.K.

    1998-02-01

    The detonation electric effect has been used to measure transit times of the MC-2453 driver subassemblies at 185 F, 212 F and room temperature after the units were subjected to a temperature of 232 F. The test procedure and the results are included in this report.

  20. Pump and Flow Control Subassembly of Thermal Control Subsystem for Photovoltaic Power Module

    NASA Technical Reports Server (NTRS)

    Motil, Brian; Santen, Mark A.

    1993-01-01

    The pump and flow control subassembly (PFCS) is an orbital replacement unit (ORU) on the Space Station Freedom photovoltaic power module (PVM). The PFCS pumps liquid ammonia at a constant rate of approximately 1170 kg/hr while providing temperature control by flow regulation between the radiator and the bypass loop. Also, housed within the ORU is an accumulator to compensate for fluid volumetric changes as well as the electronics and firmware for monitoring and control of the photovoltaic thermal control system (PVTCS). Major electronic functions include signal conditioning, data interfacing and motor control. This paper will provide a description of each major component within the PFCS along with performance test data. In addition, this paper will discuss the flow control algorithm and describe how the nickel hydrogen batteries and associated power electronics will be thermally controlled through regulation of coolant flow to the radiator.

  1. An Assessment of the International Space Station's Trace Contaminant Control Subassembly Process Economics

    NASA Technical Reports Server (NTRS)

    Perry J. L.; Cole, H. E.; El-Lessy, H. N.

    2005-01-01

    The International Space Station (ISS) Environmental Control and Life Support System includes equipment speci.cally designed to actively remove trace chemical contamination from the cabin atmosphere. In the U.S. on-orbit segment, this function is provided by the trace contaminant control subassembly (TCCS) located in the atmosphere revitalization subsystem rack housed in the laboratory module, Destiny. The TCCS employs expendable adsorbent beds to accomplish its function leading to a potentially signi.cant life cycle cost over the life of the ISS. Because maintaining the TCCSs proper can be logistically intensive, its performance in .ight has been studied in detail to determine where savings may be achieved. Details of these studies and recommendations for improving the TCCS s process economics without compromising its performance or crew health and safety are presented and discussed.

  2. Highly integrated 10Gb/s optical sub-assembly and its circuit modeling

    NASA Astrophysics Data System (ADS)

    Shim, Jongin; Kim, Dongchurl

    2006-09-01

    A highly integrated 10 Gb/s transmitter optical sub-assembly was fabricated and characterized for XFP transceiver. As a light source, uncooled 1.3 μm high-speed distributed feedback laser diode (DFB-LD) was fabricated and assembled on AlN sub-mount with a monitoring PD, a matching-resistor, and a bias-Tee with spiral-inductor. A glass sealed metallic low-loss TO-stem with in-line leads was newly presented. We developed a small-signal equivalent circuit model based on measured S-parameters in order to verify RF characteristics of LD and passive components. The eye-diagram of 10 Gb/s NRZ patterns with a PRBS 2 31 -1 was opened clearly without mask violation. At 85°C, -3-dB bandwidth was measured as high as 11 GHz and 75-km transmission was successfully achieved with very low penalty.

  3. Evaluation of Fluid Conduction and Mixing within a Subassembly of the Actinide Burner Test Reactor

    SciTech Connect

    Cliff B. Davis

    2007-09-01

    The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of the Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid, including axial and radial heat conduction and subchannel mixing, that are not currently represented with internal code models. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor.

  4. Design and performance analysis of a bio-optical sub-assembly for diffuse optical technologies

    NASA Astrophysics Data System (ADS)

    Jeong, Je-Myung; Park, Kyoungsu; Kim, Sehwan

    2014-11-01

    This paper presents a compact, multi-wavelength, and high-frequency-response light source named the bio-optical sub-assembly (BiOSA). The BiOSA is used to measure the absorption and the reduced scattering coefficients from diffuse optics-based biomedical systems. It is equipped with six laser diodes and one optical fiber with a 400- μm diameter core. Simulations can be used to determine the design parameters and to confirm the feasibility of the BiOSA. The evaluation results indicate that the coupling efficiency of the fabricated BiOSA is 80 ˜ 85%, and the frequency response is up to 3.38 GHz.

  5. Optical test bench for high precision metrology and alignment of zoom sub-assembly components

    NASA Astrophysics Data System (ADS)

    Leprêtre, F.; Levillain, E.; Wattellier, B.; Delage, P.; Brahmi, D.; Gascon, A.

    2013-09-01

    Thales Angénieux (TAGX) designs and manufactures zoom lens assemblies for cinema applications. These objectives are made of mobile lens assemblies. These need to be precisely characterized to detect alignment, polishing or glass index homogeneity errors, which amplitude may range to a few hundreds of nanometers. However these assemblies are highly aberrated with mainly spherical aberration (>30 μm PV). PHASICS and TAGX developed a solution based on the use of a PHASICS SID4HR wave front sensor. This is based on quadri-wave lateral shearing interferometry, a technology known for its high dynamic range. A 100-mm diameter He:Ne source illuminates the lens assembly entrance pupil. The transmitted wave front is then directly measured by the SID4- HR. The measured wave front (WFmeas) is then compared to a simulation from the lens sub-assembly optical design (WFdesign). We obtain a residual wave front error (WFmanufactured), which reveals lens imperfections due to its manufacturing. WFmeas=WFdesign+(WFEradius+WFEglass+WFEpolish)=WF design + WFmanufactured The optical test bench was designed so that this residual wave front is measured with a precision below 100 nm PV. The measurement of fast F-Number lenses (F/2) with aberrations up to 30 μm, with a precision of 100 nm PV was demonstrated. This bench detects mismatches in sub-assemblies before the final integration step in the zoom. Pre-alignment is also performed in order to overpass the mechanical tolerances. This facilitates the completed zoom alignment. In final, productivity gains are expected due to alignment and mounting time savings.

  6. Modeling Of Metabolic Heat Regenerated Temperature Swing Adsorption (MTSA) Subassembly For Prototype Design

    NASA Technical Reports Server (NTRS)

    Bower, Chad E.; Padilla, Sebastian A.; Iacomini, Christie S.; Paul, Heather L.

    2010-01-01

    This paper describes modeling methods for the three core components of a Metabolic heat regenerated Temperature Swing Adsorption (MTSA) subassembly: a sorbent bed, a sublimation (cooling) heat exchanger (SHX), and a condensing icing (warming) heat exchanger (CIHX). The primary function of the MTSA, removing carbon dioxide from a space suit Portable Life Support System (PLSS) ventilation loop, is performed via the sorbent bed. The CIHX is used to heat the sorbent bed for desorption and to remove moisture from the ventilation loop while the SHX is alternately employed to cool the sorbent bed via sublimation of a spray of water at low pressure to prepare the reconditioned bed for the next cycle. This paper describes subsystem heat a mass transfer modeling methodologies relevant to the description of the MTSA subassembly in Thermal Desktop and SINDA/FLUINT. Several areas of particular modeling interest are discussed. In the sorbent bed, capture of the translating carbon dioxide (CO2) front and associated local energy and mass balance in both adsorbing and desorbing modes is covered. The CIHX poses particular challenges for modeling in SINDA/FLUINT as accounting for solids states in fluid submodels are not a native capability. Methods for capturing phase change and latent heat of ice as well as the transport properties across a layer of low density accreted frost are developed. This extended modeling capacity is applicable to temperatures greater than 258 K. To extend applicability to the minimum device temperature of 235 K, a method for a mapped transformation of temperatures from below the limit temperatures to some value above is given along with descriptions for associated material property transformations and the resulting impacts to total heat and mass transfer. Similar considerations are given for the SHX along with functional relationships for areal sublimation rates as limited by flow mechanics in t1he outlet duct.

  7. Sorbent, Sublimation, and Icing Modeling Methods: Experimental Validation and Application to an Integrated MTSA Subassembly Thermal Model

    NASA Technical Reports Server (NTRS)

    Bower, Chad; Padilla, Sebastian; Iacomini, Christie; Paul, Heather L.

    2010-01-01

    This paper details the validation of modeling methods for the three core components of a Metabolic heat regenerated Temperature Swing Adsorption (MTSA) subassembly, developed for use in a Portable Life Support System (PLSS). The first core component in the subassembly is a sorbent bed, used to capture and reject metabolically produced carbon dioxide (CO2). The sorbent bed performance can be augmented with a temperature swing driven by a liquid CO2 (LCO2) sublimation heat exchanger (SHX) for cooling the sorbent bed, and a condensing, icing heat exchanger (CIHX) for warming the sorbent bed. As part of the overall MTSA effort, scaled design validation test articles for each of these three components have been independently tested in laboratory conditions. Previously described modeling methodologies developed for implementation in Thermal Desktop and SINDA/FLUINT are reviewed and updated, their application in test article models outlined, and the results of those model correlations relayed. Assessment of the applicability of each modeling methodology to the challenge of simulating the response of the test articles and their extensibility to a full scale integrated subassembly model is given. The independent verified and validated modeling methods are applied to the development of a MTSA subassembly prototype model and predictions of the subassembly performance are given. These models and modeling methodologies capture simulation of several challenging and novel physical phenomena in the Thermal Desktop and SINDA/FLUINT software suite. Novel methodologies include CO2 adsorption front tracking and associated thermal response in the sorbent bed, heat transfer associated with sublimation of entrained solid CO2 in the SHX, and water mass transfer in the form of ice as low as 210 K in the CIHX.

  8. LMFBR core design for low capital cost and low-sodium void

    SciTech Connect

    Fischer, G.J.

    1982-01-01

    The need to design LMFBR reactor cores as well as plants for lowest possible capital costs has been apparent internationally as well as in the US. At the same time it is also important to keep the sodium void reactivity gain as low as possible for safety reasons and it has always been important to assure a plant design which most effectively serves the operational needs of the utility. This paper describes a LMFBR core design which has evolved as a result of a recent effort to achieve these objectives.

  9. Review of pertinent thermal-hydraulic data for LMFBR core natural circulation analyses

    SciTech Connect

    Bishop, A. A.; Coffield, Jr., R. D.; Markley, R. A.

    1980-01-01

    A literature review and summary of significant data is presented relative to LMFBR core natural convection cooling analysis. First, a brief review of computer codes and respective input data needs is made, significant data areas are then addressed and data for verifying the code calculations are described. Recommendations and conclusions with regard to the data are included.

  10. Input parameters to codes which analyze LMFBR wire-wrapped bundles

    SciTech Connect

    Hawley, J.T.; Chan, Y.N.; Todreas, N.E.

    1980-12-01

    This report provides a current summary of recommended values of key input parameters required by ENERGY code analysis of LMFBR wire wrapped bundles. This data is based on the interpretation of experimental results from the MIT and other available laboratory programs.

  11. Process for thermal imaging scanning of a swaged heater for an anode subassembly of a hollow cathode assembly

    NASA Technical Reports Server (NTRS)

    Patterson, Michael J. (Inventor); Verhey, Timothy R. R. (Inventor); Soulas, George C. (Inventor)

    2004-01-01

    A process for thermal imaging scanning of a swaged heater of an anode subassembly of a hollow cathode assembly, comprising scanning a swaged heater with a thermal imaging radiometer to measure a temperature distribution of the heater; raising the current in a power supply to increase the temperature of the swaged heater; and measuring the swaged heater temperature using the radiometer, whereupon the temperature distribution along the length of the heater shall be less than plus or minus 5 degrees C.

  12. X-ray Digital Radiography and Computed Tomography of ICF and HEDP Materials, Subassemblies and Targets

    SciTech Connect

    Brown, W D; Martz Jr., H E

    2006-05-31

    Inertial confinement fusion (ICF) and high energy density physics (HEDP) research are being conducted at large laser facilities, such as the University of Rochester's Laboratory for Laser Energetics OMEGA facility and the Lawrence Livermore National Laboratory's (LLNL) National Ignition Facility (NIF). At such facilities, millimeter-sized targets with micrometer structures are studied in a variety of hydrodynamic, radiation transport, equation-of-state, inertial confinement fusion and high-energy density experiments. The extreme temperatures and pressures achieved in these experiments make the results susceptible to imperfections in the fabricated targets. Targets include materials varying widely in composition ({approx}3 < Z < {approx}82), density ({approx}0.03 to {approx}20 g/cm{sup 3}), geometry (planar to spherical) and embedded structures (joints to subassemblies). Fabricating these targets with structures to the tolerances required is a challenging engineering problem the ICF and HEDP community are currently undertaking. Nondestructive characterization (NDC) provides a valuable tool in material selection, component inspection, and the final pre-shot assemblies inspection. X-rays are a key method used to NDC these targets. In this paper we discuss X-ray attenuation, X-ray phase effects, and the X-ray system used, its performance and application to characterize low-temperature Raleigh-Taylor and non-cryogenic double-shell targets.

  13. Post-Flight Sampling and Loading Characterization of Trace Contaminant Control Subassembly Charcoal

    NASA Technical Reports Server (NTRS)

    Perry, J. L.; Cole, H. E.; Cramblitt, E. L.; El-Lessy, H. N.; Manuel, S.; Tucker, C. D.

    2003-01-01

    Trace chemical contaminants produced by equipment offgassing and human metabolic processes are removed from the atmosphere of the International Space Station s U.S. Segment by a trace contaminant control subassembly (TCCS). The TCCS employs a combination of physical adsorption, thermal catalytic oxidation, and chemical adsorption processes to accomplish its task. A large bed of granular activated charcoal is a primary component of the TCCS. The charcoal contained in this bed, known as the charcoal bed assembly (CBA), is expendable and must be replaced periodically. Pre-flight engineering analyses based upon TCCS performance testing results established a service life estimate of 1 year. After nearly 1 year of cumulative in-flight operations, the first CBA was returned for refurbishment. Charcoal samples were collected and analyzed for loading to determine the best estimate for the CBAs service life. A history of in-flight TCCS operations is presented as well as a discussion of the charcoal sampling procedures and chemical analysis results. A projected service life derived from the observed charcoal loading is provided. Recommendations for better managing TCCS resources are presented.

  14. Performance Testing of a Trace Contaminant Control Subassembly for the International Space Station

    NASA Technical Reports Server (NTRS)

    Perry, J. L.; Curtis, R. E.; Alexandre, K. L.; Ruggiero, L. L.; Shtessel, N.

    1998-01-01

    As part of the International Space Station (ISS) Trace Contaminant Control Subassembly (TCCS) development, a performance test has been conducted to provide reference data for flight verification analyses. This test, which used the U.S. Habitation Module (U.S. Hab) TCCS as the test article, was designed to add to the existing database on TCCS performance. Included in this database are results obtained during ISS development testing; testing of functionally similar TCCS prototype units; and bench scale testing of activated charcoal, oxidation catalyst, and granular lithium hydroxide (LiOH). The present database has served as the basis for the development and validation of a computerized TCCS process simulation model. This model serves as the primary means for verifying the ISS TCCS performance. In order to mitigate risk associated with this verification approach, the U.S. Hab TCCS performance test provides an additional set of data which serve to anchor both the process model and previously-obtained development test data to flight hardware performance. The following discussion provides relevant background followed by a summary of the test hardware, objectives, requirements, and facilities. Facility and test article performance during the test is summarized, test results are presented, and the TCCS's performance relative to past test experience is discussed. Performance predictions made with the TCCS process model are compared with the U.S. Hab TCCS test results to demonstrate its validation.

  15. Design and Assembly of an Integrated Metabolic Heat Regenerated Temperature Swing Adsorption (MTSA) Subassembly Engineering Development Unit

    NASA Technical Reports Server (NTRS)

    Padilla, Sebastian A.; Powers, Aaron; Iacomini, Christie S.; Paul, Heather L.

    2011-01-01

    Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO2) control for a Portable Life Support System (PLSS), as well as water recycling. The core of the MTSA technology is a sorbent bed that removes CO2 from the PLSS ventilation loop gas via a temperature swing. A Condensing Ice Heat eXchanger (CIHX) is used to warm the sorbent while also removing water from the ventilation loop gas. A Sublimation Heat eXchanger (SHX) is used to cool the sorbent. Research was performed to explore an MTSA designed for both lunar and Martian operations. Previously each the sorbent bed, CIHX, and SHX had been built and tested individually on a scale relevant to PLSS operations, but they had not been done so as an integrated subassembly. Design and analysis of an integrated subassembly was performed based on this prior experience and an updated transient system model. Focus was on optimizing the design for Martian operations, but the design can also be used in lunar operations. An Engineering Development Unit (EDU) of an integrated MTSA subassembly was assembled based on the design. Its fabrication is discussed. Some details on the differences between the as-assembled EDU to the future flight unit are considered.

  16. Design and Assembly of an Integrated Metabolic Heat Regenerated Temperature Swing Adsorption (MTSA) Subassembly Engineering Development Unit

    NASA Technical Reports Server (NTRS)

    Padilla, Sebastian A.; Powers, Aaron; Iacomini, Christie S.; Bower, Chad E.; Paul, Heather L.

    2012-01-01

    Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO2) control for a Portable Life Support System (PLSS), as well as water recycling. The core of the MTSA technology is a sorbent bed that removes CO2 from the PLSS ventilation loop gas via a temperature swing. A Condensing Icing Heat eXchanger (CIHX) is used to warm the sorbent while also removing water from the ventilation loop gas. A Sublimation Heat eXchanger (SHX) is used to cool the sorbent. Research was performed to explore an MTSA designed for both lunar and Martian operations. Previously the sorbent bed, CIHX, and SHX had been built and tested individually on a scale relevant to PLSS operations, but they had not been done so as an integrated subassembly. Design and analysis of an integrated subassembly was performed based on this prior experience and an updated transient system model. Focus was on optimizing the design for Martian operations, but the design can also be used in lunar operations. An Engineering Development Unit (EDU) of an integrated MTSA subassembly was assembled based on the design. Its fabrication is discussed. Some details on the differences between the as-assembled EDU and the future flight unit are considered.

  17. System Modeling of Metabolic Heat Regenerated Temperature Swing Adsorption (MTSA) Subassembly for Prototype Design

    NASA Technical Reports Server (NTRS)

    Bower, Chad; Padilla, Sebastian; Iacomini, Christie; Paul, Heather L.

    2009-01-01

    This paper describes modeling methods for the three core components of a Metabolic heat regenerated Temperature Swing Adsorption (MTSA) subassembly: the sorbent bed, a sublimation (cooling) heat exchanger (SHX), and a condensing icing (warming) heat exchanger (CIHX). The primary function of the MTSA, removing carbon dioxide from a ventilation loop, is performed via the sorbent bed. The CIHX is used to heat the sorbent bed for desorption and to remove moisture from the ventilation loop while the SHX is alternately employed to cool the sorbent bed via sublimation of a spray of water at low pressure to prepare the reconditioned bed for the next cycle. This paper describes a system level model of the MTSA as developed in Thermal Desktop and SINDA/FLUINT including assumptions on geometry and physical phenomena, modeling methodology and relevant pa ra mete rizatio ns. Several areas of particular modeling interest are discussed. In the sorbent bed, capture of the translating CO2 saturation front and associated local energy and mass balance in both adsorbing and desorbing modes is covered. The CIHX poses particular challenges for modeling in SINDA/FLUINT as accounting for solids states in fluid submodels are not a native capability. Methods for capturing phase change and latent heat of ice as well as the transport properties across a layer of low density accreted frost are developed. This extended modeling capacity is applicable to temperatures greater than 258 K. To extend applicability to the minimum device temperature of 235 K, a method for a mapped transformation of temperatures from below the limit temperatures to some value above is given along with descriptions for associated material property transformations and the resulting impacts to total heat and mass transfer. Similar considerations are shown for the SHX along with assumptions for flow mechanics and resulting model methods for sublimation in a flow.

  18. CWDM based HDMI interconnect incorporating passively aligned POF linked optical subassembly modules

    NASA Astrophysics Data System (ADS)

    Lee, Hak-Soon; Lee, Sang-Shin; Son, Yung-Sung

    2011-08-01

    A four-channel transmitter OSA (TOSA) and a receiver optical sub-assembly (ROSA) module were presented. They take advantage of a coarse WDM (CWDM) scheme, employing two types of VCSELs at 780 and 850 nm, where no wavelength filters are involved in the TOSA. The ROSA and TOSA were constructed through a fully passive alignment process using components produced by virtue of a cost effective plastic injection molding technique. In order to build a high quality optical HDMI interconnect, four channel optical links between these modules ware established via two graded-index plastic optical fibers (GI-POFs). The HDMI interconnect was thoroughly evaluated in terms of the alignment tolerance, the light beam propagation, and the data transmission capability. For the ROSA, the measured tolerance, as affected by the photodiode alignment, was ~45 μm and over 200 μm for the transverse and longitudinal directions, respectively. For the TOSA, the tolerance, which is mostly dependent upon the VCSEL alignment, was ~20 μm and more than 200 μm for the transverse and longitudinal directions, respectively. The beam profiles for the TOSA and ROSA were monitored to confirm their feasibility from the optical coupling perspective. A digital signal at 2.5 Gb/s was efficiently transmitted through the HDMI interconnect with a bit error ratio of below 10-16. A 1080p HDMI signal from a Blu-ray player was delivered through the interconnect to an LCD monitor and successfully displayed a high quality video.

  19. Final report of the APRICOT Program and results of Phase 3. [LMFBR

    SciTech Connect

    Not Available

    1982-09-01

    APRICOT (Analysis of PRImary COntainment Transients) was a cooperative activity for comparison and benchmarking of computational methods used to analyze LMFBR (Liquid Metal Fast Breeder Reactor) structural response to pressure loads from HCDA's (Hypothetical Core Disruptive Accidents). The participants were LMFBR project groups from Europe, Japan and the United States. Independent experts reviewed the calculations for the purpose of comparing computational results and methods of solution. Phase 3 involved a series of simple calculations of structural response and fluid-structure interactions under elastic and elastic-plastic conditions. The results were generally in reasonable agreement although there were a few anomalies. The APRICOT program has provided significant code validation data to enhance confidence in numerical simulations of HCDA's. It has also demonstrated the value of this type of benchmark activity.

  20. Two-dimensional modeling of sodium boiling in a simulated LMFBR loss-of-flow test

    SciTech Connect

    Rose, S.D.

    1984-01-01

    Loss-of-flow (LOF) accidents are of major importance in LMFBR safety. Tests have been performed to simulate the simultaneous failure of all primary pumps and reactor shutdown systems in a 37-pin electrically heated test bundle installed in the KNS sodium boiling loop at the Institute of Reactor Development, Karlsruhe. The tests simulated LOF conditions of the German prototype LMFBR, the SNR 300. The main objectives of these tests were to characterize the transient boiling development to cladding dryout and to provide data for validation of sodium boiling codes. One particular LOF test, designated L22, at full power was selected as a benchmark exercise for comparison of several codes at the Eleventh Meeting of the Liquid Metal Boiling Working Group (LMBWG) held in Grenoble, France, in October 1984. In this paper, the results of the calculations performed at ORNL with the two-dimensional (2-D) boiling code THORAX are presented.

  1. Numerical simulation of combined natural and forced convection during thermal-hydraulic transients. [LMFBR

    SciTech Connect

    Domanus, H.M.; Sha, W.T.

    1981-01-01

    The single-phase COMMIX (COMponent MIXing) computer code performs fully three-dimensional, transient, thermal-hydraulic analyses of liquid-sodium LMFBR components. It solves the conservation equations of mass, momentum, and energy as a boundary-value problem in space and as an initial-value problem in time. The concepts of volume porosity, surface permeability and distributed resistance, and heat source have been employed in quasi-continuum (rod-bundle) applications. Results from three transient simulations involving forced and natural convection are presented: (1) a sodium-filled horizontal pipe initially of uniform temperature undergoing an inlet velocity rundown transient, as well as an inlet temperature transient; (2) a 19-pin LMFBR rod bundle undergoing a velocity transient; and, (3) a simulation of a water test of a 1/10-scale outlet plenum undergoing both velocity and temperature transients.

  2. LMFBR system-wide transient analysis: the state of the art and US validation needs

    SciTech Connect

    Khatib-Rahbar, M.; Guppy, J.G.; Cerbone, R.J.

    1982-01-01

    This paper summarizes the computational capabilities in the area of liquid metal fast breeder reactor (LMFBR) system-wide transient analysis in the United States, identifies various numerical and physical approximations, the degree of empiricism, range of applicability, model verification and experimental needs for a wide class of protected transients, in particular, natural circulation shutdown heat removal for both loop- and pool-type plants.

  3. LMFBR conceptual design study: an overview of environmental and safety concerns

    SciTech Connect

    Brenchley, D.L.

    1981-06-01

    The US Department of Energy (DOE) initiated the Liquid Metal Fast Breeder (LMFBR) Conceptual Design Study (CDS) with the objective of maintaining a viable breeder option. The project is scheduled to be completed in FY-1981 but decisions regarding plant construction will be delayed until at least 1985. This report provides a review of the potential environmental and safety engineering concerns for the CDS and recommends specific action for the Environmental and Safety Engineering Division of DOE.

  4. PFR/TREAT program: objectives, accomplishments, and plans. [LMFBR

    SciTech Connect

    Cowking, C.B.; Alter, H.; Stillwell, J.; Wood, M.H.; Woods, W.J.; Culley, G.E.; Klickman, A.E.; Borys, S.S.

    1984-01-01

    The PFR-TREAT collaborative program of transient safety testing of fast reactor fuel was established in 1979 to provide mutual advantage to USDOE and the UKAEA through irradiation of US and UK full-length fuel pins in PFR, followed by safety testing in TREAT. The tests which were planned include Transient Over-Power (TOP) and Transient Under-Cooling with Over-Power (TUCOP) tests to fuel destruction and re-distribution; the results will provide significant new information on fuel and cladding behavior in hypothetical reactor faults. The information obtained in both US and UK fuel pins is to be interpreted by both partners and published jointly when mutually agreed. Thirteen tests, on fresh and irradiated fuel, in single-pin and 7-pin test sections, were completed by the end of 1983. The test matrix, which is currently being re-evaluated, calls for additional tests to be run under the present agreement. There has been an extensive program of post irradiation examination of sibling pins in both the UK and the US to characterize the test fuel prior to destructive irradiation, including testing of irradiated cladding to determine its failure characteristics.

  5. US LMFBR (Liquid Metal Fast Breeder Reactor): flow induced vibration program (1977-1986): A summary and overview

    SciTech Connect

    Wambsganss, M.W.; Chen, S.S.; Mulcahy, T.M.; Jendrzejczyk, J.A.

    1986-09-01

    This paper summarizes the activities and accomplishments under the US LMFBR Flow Induced Vibration Program for the period 1977-1986. Since 1977 represents the date of the last IAEA IWGFR Specialists Meeting on LMFBR Flow Induced Vibration, this paper thus provides an update to the results presented at that meeting. This period also represents a period of substantial change for the US LMFBR program. A major reactor project, the FFTF, was completed and a second major project, the CRBR plant, was terminated. This change adversely impacted the US flow induced vibration program. Nevertheless, base technology activities have continued. In this paper, research in the following areas is summarized: Vibration characteristics and scaling, Turbulent buffeting and vortex shedding, Fluidelastic instabilities of tube bundles in crossflow, and Instabilities induced by leakage flows.

  6. SSME Alternate Turbopump Development Program: Design verification specification for high-pressure fuel turbopump

    NASA Technical Reports Server (NTRS)

    1989-01-01

    The design and verification requirements are defined which are appropriate to hardware at the detail, subassembly, component, and engine levels and to correlate these requirements to the development demonstrations which provides verification that design objectives are achieved. The high pressure fuel turbopump requirements verification matrix provides correlation between design requirements and the tests required to verify that the requirement have been met.

  7. Novel bidirectional optical subassembly with embedded filter, 45-degree angle polished fiber cladding and etched fiber core

    NASA Astrophysics Data System (ADS)

    Lee, Seihyoung; Lim, Kwon-Seob; Lee, Jong Jin; Kang, Hyun Seo

    2009-10-01

    The optical wavelength-division-multiplex filter for bidirectional optical subassembly (BOSA) is embedded to the fiber core, which results in simplicity of the BOSA module. The fiber cladding is 45-deg angle polished to receive a downstream signal. The core is etched by a femtosecond laser to have a normal core facet and to transmit an upstream signal. The downstream signal, which is core mode, is coupled to the cladding mode by the long-period fiber grating and then detected by a photodiode by means of the total internal reflection effect at the 45-deg angle polished cladding facet. The measured transmitted and received coupling efficiencies are 27.3 and 43.8%, respectively.

  8. Electrospray ionization mass spectrometric determination of the molecular mass of the approximately 200-kDa globin dodecamer subassemblies in hexagonal bilayer hemoglobins.

    PubMed

    Green, B N; Bordoli, R S; Hanin, L G; Lallier, F H; Toulmond, A; Vinogradov, S N

    1999-10-01

    Hexagonal bilayer hemoglobins (Hbs) are approximately 3.6-MDa complexes of approximately 17-kDa globin chains and 24-32-kDa, nonglobin linker chains in a approximately 2:1 mass ratio found in annelids and related species. Studies of the dissociation and reassembly of Lumbricus terrestris Hb have provided ample evidence for the presence of a approximately 200-kDa linker-free subassembly consisting of monomer (M) and disulfide-bonded trimer (T) subunits. Electrospray ionization mass spectrometry (ESI-MS) of the subassemblies obtained by gel filtration of partially dissociated L. terrestris and Arenicola marina Hbs showed the presence of noncovalent complexes of M and T subunits with masses in the 213. 3-215.4 and 204.6-205.6 kDa ranges, respectively. The observed mass of the L. terrestris subassembly decreased linearly with an increase in de-clustering voltage from approximately 215,400 Da at 60 V to approximately 213,300 Da at 200 V. In contrast, the mass of the A. marina complex decreased linearly from 60 to 120 V and reached an asymptote at approximately 204,600 Da (180-200 V). The decrease in mass was probably due to the progressive removal of complexed water and alkali metal cations. ESI-MS at an acidic pH showed both subassemblies to consist of only M and T subunits, and the experimental masses demonstrated them to have the composition M(3)T(3). Because there are three isoforms of M and four isoforms of T in Lumbricus and two isoforms of M and 5 isoforms of T in Arenicola, the masses of the M(3)T(3) subassemblies are not unique. A random assembly model was used to calculate the mass distributions of the subassemblies, using the known ESI-MS masses and relative intensities of the M and T subunit isforms. The expected mass of randomly assembled subassemblies was 213,436 Da for Lumbricus Hb and 204,342 Da for Arenicola Hb, in good agreement with the experimental values. PMID:10497174

  9. 85,000-GPM, single-stage, single-suction LMFBR intermediate centrifugal pump

    SciTech Connect

    Fair, C.E.; Cook, M.E. Huber, K.A.; Rohde, R.

    1983-01-01

    The mechanical and hydraulic design features of the 85,000-gpm, single-stage, single-suction pump test article, which is designed to circulate liquid-sodium coolant in the intermediate heat-transport system of a Large-Scale Liquid Metal Fast Breeder Reactor (LS-LMFBR), are described. The design and analytical considerations used to satisfy the pump performance and operability requirements are presented. The validation of pump hydraulic performance using a hydraulic scale-model pump is discussed, as is the featute test for the mechanical-shaft seal system.

  10. Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

    SciTech Connect

    Serell, D.C.; Kaplan, S.

    1980-09-01

    Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers.

  11. 4 channel × 10 Gb/s bidirectional optical subassembly using silicon optical bench with precise passive optical alignment.

    PubMed

    Kang, Eun Kyu; Lee, Yong Woo; Ravindran, Sooraj; Lee, Jun Ki; Choi, Hee Ju; Ju, Gun Wu; Min, Jung Wook; Song, Young Min; Sohn, Ik-Bu; Lee, Yong Tak

    2016-05-16

    We demonstrate an advanced structure for optical interconnect consisting of 4 channel × 10 Gb/s bidirectional optical subassembly (BOSA) formed using silicon optical bench (SiOB) with tapered fiber guiding holes (TFGHs) for precise and passive optical alignment of vertical-cavity surface-emitting laser (VCSEL)-to-multi mode fiber (MMF) and MMF-to-photodiode (PD). The co-planar waveguide (CPW) transmission line (Tline) was formed on the backside of silicon substrate to reduce the insertion loss of electrical data signal. The 4 channel VCSEL and PD array are attached at the end of CPW Tline using a flip-chip bonder and solder pad. The 12-channel ribbon fiber is simply inserted into the TFGHs of SiOB and is passively aligned to the VCSEL and PD in which no additional coupling optics are required. The fabricated BOSA shows high coupling efficiency and good performance with the clearly open eye patterns and a very low bit error rate of less than 10-12 order at a data rate of 10 Gb/s with a PRBS pattern of 231-1. PMID:27409898

  12. Fuel Storage Facility Final Safety Analysis Report. Revision 1

    SciTech Connect

    Linderoth, C.E.

    1984-03-01

    The Fuel Storage Facility (FSF) is an integral part of the Fast Flux Test Facility. Its purpose is to provide long-term storage (20-year design life) for spent fuel core elements used to provide the fast flux environment in FFTF, and for test fuel pins, components and subassemblies that have been irradiated in the fast flux environment. This Final Safety Analysis Report (FSAR) and its supporting documentation provides a complete description and safety evaluation of the site, the plant design, operations, and potential accidents.

  13. Results of phase 2 of the APRICOT program. Final report. [LMFBR

    SciTech Connect

    Not Available

    1981-05-01

    APRICOT (Analysis of Primary Containment Transients) is a cooperative activity for comparison and benchmarking of computational methods used to analyze LMFBR (Liquid Metal Fast Breeder Reactor) structural response to pressure loads from HCDA's (Hypothetical Core Disruptive Accidents). Independent experts review the calculations for the purpose of comparing computational results and methods of solution. Phase 2 involved a series of more complex calculations based on the simulation of scaled-down containment experiments. These calculations, as those of Phase 1, were performed by participants from Europe, Japan and the United States. The calculations were all in reasonable agreement with experimental determinations of hydrodynamic loads; however, the calculated plastic strains differed significantly from the experimental results. The unresolved issues from the Phase 2 calculations are currently being studied with the calculations for Phase 3.

  14. Mitigation of thermal transients by tube bundle inlet plenum design. [LMFBR

    SciTech Connect

    Oras, J.J.; Kasza, K.E.

    1984-06-01

    A multiphase program aimed at investigating the importance of thermal buoyancy to LMFBR steam-generator and heat-exchanger thermal hydraulics under low-flow transient conditions is being conducted in the Argonne Mixing Components Test Facility (MCTF) on a 60/sup 0/ sector shell-side flow model of the Westinghouse straight-tube steam generator being developed under the US/DOE large-component development program. A series of shell-side constant-flow thermal-downramp transient tests have been conducted focusing on the phenomenon of thermal-buoyancy-induced-flow channeling. In addition, it was discovered that a shell-inlet flow-distribution plenum can play a significant role in mitigating the severity of a thermal transient entering a steam generator or heat exchanger.

  15. Finite-element blunt-crack propagation: a modified J-integral approach. [LMFBR

    SciTech Connect

    Pan, Y.C.; Marchertas, A.H.; Kennedy, J.M.

    1983-01-01

    In assessing the safety of a liquid metal fast breeder reactor (LMFBR), a major concern is the behavior of concrete structures subjected to high temperatures. The potential of concrete cracking is an important parameter which could significantly influence the safety assessment of thermally attacked concrete. A new modified J-integral approach for the blunt crack model has been derived to provide a general procedure to accurately predict the direction of crack growth. This formulation has been incorporated into the coupled heat transfer-stress analysis finite element code TEMP-STRESS. A description of the formulation is presented in this paper. Results for the problems of a Mode I and mixed mode crack in a plate using regular and slanted meshes subjected to uniaxial and shear loading are presented.

  16. DYNAPCON: a computer code for dynamic analysis of prestressed concrete structures. [LMFBR

    SciTech Connect

    Marchertas, A.H.

    1982-09-01

    A finite element computer code for the transient analysis of prestressed concrete reactor vessels (PCRVs) for LMFBR containment is described. The method assumes rotational symmetry of the structure. Time integration is by an explicit method. The quasistatic prestressing operation of the PCRV model is performed by a dynamic relaxation technique. The material model accounts for the crushing and tensile cracking in arbitrary direction in concrete and the elastic-plastic behavior of reinforcing steel. The variation of the concrete tensile cracking and compressive crushing limits with strain rate is taken into account. Relative slip is permitted between the concrete and tendons. Several example solutions are presented and compared with experimental results. These sample problems range from simply supported beams to small scale models of PCRV's. It is shown that the analytical methods correlate quite well with experimental results, although in the vicinity of the failure load the response of the models tend to be quite sensitive to input parameters.

  17. System seismic analysis of an innovative primary system for a large pool type LMFBR plant

    SciTech Connect

    Pan, Y.C.; Wu, T.S.; Cha, B.K.; Burelbach, J.; Seidensticker, R.

    1984-01-01

    The system seismic analysis of an innovative primary system for a large pool type liquid metal fast breeder reactor (LMFBR) plant is presented. In this primary system, the reactor core is supported in a way which differs significantly from that used in previous designs. The analytical model developed for this study is a three-dimensional finite element model including one-half of the primary system cut along the plane of symmetry. The model includes the deck and deck mounted components,the reactor vessel, the core support structure, the core barrel, the radial neutron shield, the redan, and the conical support skirt. The sodium contained in the primary system is treated as a lumped mass appropriately distributed among various components. The significant seismic behavior as well as the advantages of this primary system design are discussed in detail.

  18. Assessment of inspectability of LMFBR designs. Final report. [1000 MW(e)

    SciTech Connect

    Not Available

    1981-09-01

    This two-volume report provides a comprehensive review of the inspectability of specific portions of loop- and pool-type LMFBR (1000-MWe) designs selected by EPRI. The designs were developed during the mid to late 1970s by three independent design teams (General Electric Co., Rockwell International, and Westinghouse) under the sponsorship of DOE (formerly ERDA) and EPRI. The requirements for normal, contingency, and post-repair inspections, addressed in this report, were established from Draft 12 of the ASME Boiler and Pressure Vessel Code, Section XI Division 3, issued in September 1979. These requirements, the intrinsic characteristics of the designs, the environmental (radiation, thermal, and atmospheric) aspects, and the available (present and near-term) inspection techniques, formed the basis for assessing the selected portions of the design or (1) accessibility, (2) feasibility, (3) practicality, and (4) costs to perform the above-specified inspections.

  19. Temperature Distributions in LMR Fuel Pin Bundles as Modeled by COBRA-IV-I

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Stout, Sherry

    2005-02-01

    Most pin type reactor designs for space power or terrestrial applications group the fuel pins into a number of relatively large fuel pin bundles or subassemblies. Fuel bundles for terrestrial liquid metal fast breeders reactors typically use 217 - 271 pins per sub-assembly, while some SP100 designs use up to 331 pins in a central subassembly that was surrounded by partial assemblies. Because thermal creep is exponentially related to temperature, small changes in fuel pin cladding temperature can make large differences in the lifetime in a high temperature liquid metal reactor (LMR). This paper uses the COBRA-IV-I computer code to determine the temperature distribution within LMR fuel bundles. COBRA-IV-I uses the sub-channel analysis approach to determine the enthalpy (or temperature) and flow distribution in rod bundles for both steady-state and transient conditions. The COBRA code runs in only a few seconds and has been benchmarked and tested extensively over a wide range of flow conditions. In this report the flow and temperature distributions for two types of lithium cooled space reactor core designs were calculated. One design uses a very tight fuel pin packing that has a pitch to diameter ratio of 1.05 (small wire wrap with a diameter of 392 μm) as proposed in SP100. The other design uses a larger pitch to diameter ratio of 1.09 with a larger more conventional sized wire wrap diameter of 1 mm. The results of the COBRA pin bundle calculations show that the larger pitch-to-diameter fuel bundle designs are more tolerant to local flow blockages, and in addition they are less sensitive to mal-flow distributions that occur near the edges of the subassembly.

  20. Temperature Distributions in LMR Fuel Pin Bundles as Modeled by COBRA-IV-I

    SciTech Connect

    Wright, Steven A.; Stout, Sherry

    2005-02-06

    Most pin type reactor designs for space power or terrestrial applications group the fuel pins into a number of relatively large fuel pin bundles or subassemblies. Fuel bundles for terrestrial liquid metal fast breeders reactors typically use 217 - 271 pins per sub-assembly, while some SP100 designs use up to 331 pins in a central subassembly that was surrounded by partial assemblies. Because thermal creep is exponentially related to temperature, small changes in fuel pin cladding temperature can make large differences in the lifetime in a high temperature liquid metal reactor (LMR). This paper uses the COBRA-IV-I computer code to determine the temperature distribution within LMR fuel bundles. COBRA-IV-I uses the sub-channel analysis approach to determine the enthalpy (or temperature) and flow distribution in rod bundles for both steady-state and transient conditions. The COBRA code runs in only a few seconds and has been benchmarked and tested extensively over a wide range of flow conditions. In this report the flow and temperature distributions for two types of lithium cooled space reactor core designs were calculated. One design uses a very tight fuel pin packing that has a pitch to diameter ratio of 1.05 (small wire wrap with a diameter of 392 {mu}m) as proposed in SP100. The other design uses a larger pitch to diameter ratio of 1.09 with a larger more conventional sized wire wrap diameter of 1 mm. The results of the COBRA pin bundle calculations show that the larger pitch-to-diameter fuel bundle designs are more tolerant to local flow blockages, and in addition they are less sensitive to mal-flow distributions that occur near the edges of the subassembly.

  1. LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

    SciTech Connect

    Khan, E.U.

    1980-04-01

    A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist.

  2. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) heat transport system dynamics and steam generator control: Figures

    NASA Astrophysics Data System (ADS)

    Brukx, J. F. L. M.

    1982-06-01

    Dynamic modeling of LMFBR heat transport system is discussed. Uncontrolled transient behavior of individual components and of the integrated heat transport system are considered. For each component, results showing specific dynamic features of the component and/or model capability were generated. Controlled dynamic behavior for alternative steam generator control systems during forced and natural sodium coolant circulation was analyzed. Combined free and forced convection of laminar and turbulent vertical pipe flow of liquid metals was investigated.

  3. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) neat transport system dynamics and steam generator control

    NASA Astrophysics Data System (ADS)

    Brukx, J. F. L. M.

    1982-06-01

    Loop type LMFBR heat transport system dynamics after reactor shutdown and during subsequent decay heat removal are considered with emphasis on steam generator dynamics including the development and evaluation of various post-scram steam generator control systems, and natural circulation of the sodium coolant, including the influence of superimposed free convection on forced convection heat transfer and pressure drop. The normal operating and decay heat removal functions of the overall heat transport system are described.

  4. Influence of thermal buoyancy on vertical tube bundle thermal density head predictions under transient conditions. [LMFBR

    SciTech Connect

    Lin, H.C.; Kasza, K.E.

    1984-01-01

    The thermal-hydraulic behavior of an LMFBR system under various types of plant transients is usually studied using one-dimensional (1-D) flow and energy transport models of the system components. Many of the transient events involve the change from a high to a low flow with an accompanying change in temperature of the fluid passing through the components which can be conductive to significant thermal bouyancy forces. Thermal bouyancy can exert its influence on system dynamic energy transport predictions through alterations of flow and thermal distributions which in turn can influence decay heat removal, system-response time constants, heat transport between primary and secondary systems, and thermal energy rejection at the reactor heat sink, i.e., the steam generator. In this paper the results from a comparison of a 1-D model prediction and experimental data for vertical tube bundle overall thermal density head and outlet temperature under transient conditions causing varying degrees of thermal bouyancy are presented. These comparisons are being used to generate insight into how, when, and to what degree thermal buoyancy can cause departures from 1-D model predictions.

  5. Progress report on LLTR Series II Test A-2 (Part 1). [LMFBR

    SciTech Connect

    Freede, W.J.; Neely, H.H.

    1980-01-01

    This document contains a complete set of valid and final digital and analog data plots for LLTR Series II, Test A-2. Included is an Accuracy Statement regarding this data as required by Revision 0 of the GE Test Request, Specification No. 23A2062. The Series II, Sodium-Water Reaction Test A-2 was performed in the Large Leak Test Rig (LLTR) at the Energy Technology Engineering Center (ETEC). This was the third of three planned double-edged guillotine (DEG) rupture tests of a single tube which will be followed by a number of small leak tests. The test article is the LLTI which is a full-size diameter internals, shortened in length and prototypic of the CRBR steam generator. It is installed in the Large Leak Test Vessel (LLTV). The overall test program was formulated by General Electric (GE) as Test Requester to establish steam generator design and to verify analytical models/codes to estimate the effect of large leak accidents in an LMFBR demonstration plant steam generator and system.

  6. Simulation of LMFBR pump transients and comparison to LOF that occurred at EBR-II

    SciTech Connect

    Koenig, F.F.; Dean, E.M.

    1985-01-01

    In a large LMFBR plant design, a number of pumps in parallel will feed the core. It must be demonstrated that the plant can continue to operate with the loss of one of the primary pumps. It is desirable not to have check valves in the loop from a reliability and economic standpoint. Simulations have been made to determine the consequences of a loss of one pump in a four-loop pool plant in which no plant protection action is taken. This analysis would be used to determine the required power rundown that would accompany pump loss. The two primary centrifugal pumps in EBR-II feed the core and blanket plenums in two parallel flow paths. The loss of one pump will result in decrease core flow and reverse flow through the down pump since no check valves are present in the system. For a large pool plant with four primary pumps, the loss of one pump will also result in reverse flow through the down pump if check valves of flow diodes are not included. The resulting flow transient has been modeled for EBR-II and the large plant using the DNSP program.

  7. Observation of large, non-covalent globin subassemblies in the approximately 3600 kDa hexagonal bilayer hemoglobins by electrospray ionization time-of-flight mass spectrometry.

    PubMed

    Green, B N; Gotoh, T; Suzuki, T; Zal, F; Lallier, F H; Toulmond, A; Vinogradov, S N

    2001-06-01

    A non-covalent globin subassembly comprising 12 globin chains (204 to 214 kDa) was observed directly by electrospray ionization time-of-flight mass spectrometry in the native hexagonal bilayer hemoglobins from the oligochaetes Lumbricus terrestris and Tubifex tubifex, the polychaetes Tylorrhynchus heterochaetus, Arenicola marina, Amphitrite ornata and Alvinella pompejana, the leeches Macrobdella decora, Haemopis grandis and Nephelopsis oscura and the chlorocruorin from the polychaete Myxicola infundibulum, over the pH range 3.5-7.0. The Hb from the deep-sea polychaete Alvinella exhibited in addition, peaks at approximately 107 kDa and at approximately 285 kDa, which were assigned to subassemblies of six globin chains and of 12 globin chains with three non-globin linker chains, respectively. The experimental masses decreased slightly with increased de-clustering potential (60 to 160 V) and were generally 0.1 to 0.2 % higher than the calculated masses, due probably to complexation with cations and water molecules. PMID:11397079

  8. Bidirectional optical subassembly-shaped 20-Gbit/s compact single-mode four-channel wavelength-division multiplexing optical modules for optical multimedia interfaces

    NASA Astrophysics Data System (ADS)

    Lim, Kwon-Seob; Yu, Hong-Yeon; Park, Hyoung-Jun; Kang, Hyun Seo; Jang, Jae-Hyung

    2016-06-01

    Low-cost single-mode four-channel optical transmitter and receiver modules using the wavelength-division multiplexing (WDM) method have been developed for long-reach fiber optic applications. The single-mode four-channel WDM optical transmitter and receiver modules consist of two dual-wavelength optical transmitter and receiver submodules, respectively. The integration of two channels in a glass-sealed transistor outline-can package is an effective way to reduce cost and size and to extend the number of channels. The clear eye diagrams with more than about 6 dB of the extinction ratio and the minimum receiver sensitivity of lower than -16 dBm at a bit error rate of 10-12 have been obtained for the transmitter and receiver modules, respectively, at 5 Gbps/channel. The 4K ultrahigh definition contents have been transmitted over a 1-km-long single-mode fiber using a pair of proposed four-channel transmitter optical subassembly and receiver optical subassembly.

  9. Fission gas release from oxide fuels at high burnups (AWBA development program)

    SciTech Connect

    Dollins, C.C.

    1981-02-01

    The steady state gas release, swelling and densification model previously developed for oxide fuels has been modified to accommodate the slow transients in temperature, temperature gradient, fission rate and pressure that are encountered in normal reactor operation. The gas release predictions made by the model were then compared to gas release data on LMFBR-EBRII fuels obtained by Dutt and Baker and reported by Meyer, Beyer, and Voglewede. Good agreement between the model and the data was found. A comparison between the model and three other sets of gas release data is also shown, again with good agreement.

  10. Theory and use of GIRAFFE for analysis of decay characteristics of delayed-neutron precursors in an LMFBR

    SciTech Connect

    Gross, K. C.

    1980-07-01

    The application of the computer code GIRAFFE (General Isotope Release Analysis For Failed Elements) written in FORTRAN IV is described. GIRAFFE was designed to provide parameter estimates of the nonlinear discrete-measurement models that govern the transport and decay of delayed-neutron precursors in a liquid-metal fast breeder reactor (LMFBR). The code has been organized into a set of small, relatively independent and well-defined modules to facilitate modification and maintenance. The program logic, the numerical techniques, and the methods of solution used by the code are presented, and the functions of the MAIN program and of each subroutine are discussed.

  11. Development of models for the two-dimensional, two-fluid code for sodium boiling NATOF-2D. [LMFBR

    SciTech Connect

    Zielinski, R.G.; Kazimi, M.S.

    1981-09-01

    Several features were incorporated into NATOF-2D, a two-dimensional, two fluid code developed at MIT for the purpose of analysis of sodium boiling transients under LMFBR conditions. They include improved interfacial mass, momentum and energy exchange rate models, and a cell-to-cell radial heat conduction mechanism which was calibrated by simulation of Westinghouse Blanket Heat Transfer Test Program Runs 544 and 545. Finally, a direct method of pressure field solution was implemented into a direct method of pressure field solution was implemented into NATOF-2D, replacing the iterative technique previously available, and resulted in substantially reduced computational costs.

  12. Status of advanced fuel candidates for Sodium Fast Reactor within the Generation IV International Forum

    SciTech Connect

    F. Delage; J. Carmack; C. B. Lee; T. Mizuno; M. Pelletier; J. Somers

    2013-10-01

    The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxide and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.

  13. Environmental Assessment for DOE permission for off-loading activities to support the movement of Millstone Unit 2 steam generator sub-assemblies across the Savannah River Site

    SciTech Connect

    Not Available

    1992-10-01

    The Department of Energy (DOE) has prepared an Environmental Assessment (EA), for the proposed granting of DOE permission of offloading activities to support the movement Millstone Unit 2 steam generator sub-assemblies (SGSAs) across the Savannah River Site (SRS). Based on the analyses in the EA, DOE has determined that the proposed action is not 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, and the Department is issuing this Finding of No Significant Impact. On the basis of the floodplain/wetlands assessment in the EA, DOE has determined that there is no practicable alternative to the proposed activities and that the proposed action has been designed to minimize potential harm to or within the floodplain of the SRS boat ramp. No wetlands on SRS would be affected by the proposed action.

  14. R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

    SciTech Connect

    Vanderhaegen, M.; Paumel, K.; Seiler, J. M.; Tourin, A.; Jeannot, J. P.; Rodriguez, G.

    2011-07-01

    In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boiling intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)

  15. Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

    SciTech Connect

    Henderson, J M; Wood, S A; Knight, D D

    1981-01-01

    The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting.

  16. Loss-of-flow transient characterization in carbide-fueled LMFBRs

    NASA Astrophysics Data System (ADS)

    Rothrock, R. B.; Morgan, M. M.; Baars, R. E.; Elson, J. S.; Wray, M. L.

    One of the benefits derived from the use of carbide fuel in advanced liquid metal fast breeder reactors (LMFBRs) is a decreased vulnerability to certaiin accidents. This can be achieved through the combination of advanced fuel performance with the enhanced reactivity feedback effects and passive shutdown cooling systems characteristic of the current inherently safe plant concepts. The calculated core response to an unprotected loss of flow (ULOF) accident has frequently been used as a benchmark test of these designs, and the advantages of a high-conductivity fuel in relation to this type of transient have been noted in previous analyses. To evaluate this benefit in carbide-fueled LMFBRs incorporating representative current plant design features, limited calculations have been made of a ULOF transient in a small (modular) carbide-fueled LMFBR.

  17. THORAX pretest prediction of a sodium-boiling transient in a 19-pin simulated LMFBR driver bundle

    SciTech Connect

    Rose, S.D.

    1982-01-01

    Experiments will be conducted in the Thermal-Hydraulic Out-of-Reactor Safety-Shutdown Heat Removal System (THORS-SHRS) Assembly 1 loop at Oak Ridge National Laboratory (ORNL) to model the behavior of a reactor during degraded decay heat removal conditions. The test section is to consist of two parallel 19-pin electrically-heated driver bundles, typical of U.S. Large Developmental Plant (LDP) Liquid Metal Fast Breeder Reactor (LMFBR) design. Analysis of these experiments will include using THORAX, a two-dimensional boiling model which assumes an equilibrium mixture two-phase flow (with slip). A THORAX prediction is presented for a single-bundle forced convection boiling-to-dryout transient at 15.8 kW/pin.

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

    SciTech Connect

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

    1983-02-01

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

  19. Breeder Spent Fuel Handling (BSFH) cask study for FY83. Final report

    SciTech Connect

    Diggs, J M

    1985-01-01

    This report documents a study conducted to investigate the applicability of existing LWR casks to shipment of long-cooled LMFBR fuel from the Clinch River Breeder Reactor Plant (CRBRP) to the Breeder Reprocessing Engineering Test (BRET) Facility. This study considered a base case of physical constraints of plants and casks, handling capabilities of plants, through-put requirements, shielding requirements due to transportation regulation, and heat transfer capabilities of the cask designs. Each cask design was measured relative to the base case. 15 references, 4 figures, 6 tables.

  20. Update on the Hot Fuel Examination Facility (HFEF) complex

    SciTech Connect

    Bacca, J.P.

    1985-01-01

    The Hot Fuel Examination Facility (HFEF) Complex continues to provide primary postirradiation handling and examination support for the US, US/Japanese, and US/United Kingdom Liquid Metal Fast Breeder Reactor (LMFBR) fuels and materials development and reactor safety programs. For the last several years, such support has also been provided for other US nuclear programs. These include the Light Water Reactor Safety Program, the Three Mile Island No. 2 (TMI-2) reactor accident-recovery programs, the Light Water Breeder Reactor Program, the US Air Force Neutron Radiography Program, and Defense Nuclear Programs. The HFEF facilities have been refurbished and upgraded and capabilities have been added to accommodate these programs and to maintain the HFEF Complex as a world-class, state-of-the-art hot-cell complex.

  1. Fuel flexible fuel injector

    SciTech Connect

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  2. W-1 SLSF post-test data analysis. Part 1. Thermal hydraulic analysis. [LMFBR

    SciTech Connect

    Knight, D.D.

    1980-10-01

    Four types of tests were performed: (1) a decay heat transient test, (2) Loss-of-Piping-Integrity (LOPI) tests, (3) Boiling Window Tests (BWT), and (4) a fuel pin dryout and failure test. In addition, preliminary tests were run to check systems performance, instrumentation performance and test section heat balance. The objective of the decay heat test was to determine the decay heat transfer characteristics of fresh fuel pins with subcooled sodium. The objective of the LOPI experiments was to test the thermal behavior of fuel pins with four different fuel conditions subjected to the same transient. The transient was designed to simulate a rapid flow decrease as a result of pipe rupture followed by a reactor scram. The objective of the Boiling Window Tests was to study boiling initiation and progression of boiling within the fuel pin bundle.

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

    SciTech Connect

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

    1982-08-01

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

  4. Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

    NASA Astrophysics Data System (ADS)

    Seiler, J. M.; Rameau, B.

    Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

  5. HEDL W-1 SLSF experiment LOPI transient and boiling test results. [LMFBR

    SciTech Connect

    Henderson, J.M.; Wood, S.A.; Rothrock, R.B.

    1980-01-01

    The W-1 Sodium Loop Safety Facility (SLSF) experiment was designed to study the heat release characteristics of fast reactor fuel pins under Loss-of-Piping-Integrity (LOPI) accident conditions and determine stable sodium boiling initiation and recovery limits in a prototypic fuel pin bundle array. The results of the experiment address major second level of assurance (LOA-2) safety issues and provide increased insight and understanding of phenomena that would inherently terminate hypothesized accidents with only limited core damage. The irradiation phase of the experiment, consisting of thirteen individual transients, was performed between May 27 and July 20, 1979. The final transient produced approximately two seconds of coolant boiling, cladding dryout, and incipient fuel pin failure. The facility and test hardware performed as designed, allowing completion of all planned tests and achievement of all test objectives.

  6. Fabrication of UO/sub 2/-stainless steel fuel for a fast-reactor safety test

    SciTech Connect

    Rhude, H.V.; O'Keefe, G.B.; Noland, R.A.

    1983-01-01

    As part of its fast breeder reactor (LMFBR) safety research program, Argonne National Laboratory is conducting a series of tests in the TREAT Reactor to clarify the boilup and freezing behavior of UO/sub 2/ fuel and stainless steel in the transition phase of an accident sequence. Although much work has been done on UO/sub 2/ dispersed in a stainless steel matrix, no references could be found on work relating to the dispersal of stainless steel in a UO/sub 2/ matrix. Consequently a development program was started to produce the necessary fuel. The major difficulty encountered is attributable to the melting range of 316 stainless steel (1391 to 1399/sup 0/C) which is well below the normal sintering temperature of UO/sub 2/ (1750/sup 0/C). Attempts to produce the required fuel pellets by standard sintering techniques did not work. Alternative approaches to sintering were investigated to determine the density achievable without sintering.

  7. Comparisons of inelastic J and J* evaluations for the blunt crack and the sharp crack models. [LMFBR

    SciTech Connect

    Pan, Y.C.; Marchertas, A.H.; Kennedy, J.M.

    1984-01-01

    Concrete cracking is an important consideration in assessing the safety of a liquid metal fast breeder reactor (LMFBR) plant under a hypothetical accident where molten metal may come into contact with concrete structures. At the present time, several options in modeling concrete cracking have been pursued in an ongoing research program at Argonne National Laboratory which encompasses many aspects of high temperature behavior of concrete. Main emphasis is currently given to the blunt crack model where the crack is assumed to be uniformly distributed throughout the area of an element, though the sharp crack model is still kept as an alternative option where the crack surface is treated as the boundary of the finite element mesh. Several crack propagation criteria have been considered. Among these is the development of the J-integral approach with the blunt crack model. Numerical results were compared with those of the sharp crack model and found to be in good agreement for the elastic problem of a mode I crack. In this paper, the J-integral approach is extended to the post yield regime. To examine the path independency, the J* integral option is added to the finite element code. Numerical results for the J and J* integral formulations are given for a three point bending specimen loaded beyond the yield point.

  8. DIF3D nodal neutronics option for two- and three-dimensional diffusion theory calculations in hexagonal geometry. [LMFBR

    SciTech Connect

    Lawrence, R.D.

    1983-03-01

    A nodal method is developed for the solution of the neutron-diffusion equation in two- and three-dimensional hexagonal geometries. The nodal scheme has been incorporated as an option in the finite-difference diffusion-theory code DIF3D, and is intended for use in the analysis of current LMFBR designs. The nodal equations are derived using higher-order polynomial approximations to the spatial dependence of the flux within the hexagonal-z node. The final equations, which are cast in the form of inhomogeneous response-matrix equations for each energy group, involved spatial moments of the node-interior flux distribution plus surface-averaged partial currents across the faces of the node. These equations are solved using a conventional fission-source iteration accelerated by coarse-mesh rebalance and asymptotic source extrapolation. This report describes the mathematical development and numerical solution of the nodal equations, as well as the use of the nodal option and details concerning its programming structure. This latter information is intended to supplement the information provided in the separate documentation of the DIF3D code.

  9. D10 experiment: coolability of UO/sub 2/ debris in sodium with downward heat removal. [LMFBR

    SciTech Connect

    Mitchell, G.W.; Ottinger, C.A.; Meister, H.

    1984-12-01

    The LMFBR Debris Coolability Program at Sandia National Laboratories investigates the coolability of particle beds which may form following a severe accident involving core disassembly in a nuclear reactor. The D series experiments utilize fission heating of fully enriched UO/sub 2/ particles submerged in sodium to realistically simulate decay heating. The D10 experiment is the first in the series to study the effects of bottom cooling of the debris that could be provided in an actual accident condition by structural materials onto which the debris might settle. Additionally, the D10 experiment was designed to achieve maximum temperatures in the debris approaching the melting point of UO/sub 2/. The experiment was successfully operated for over 50 hours and investigated downward heat removal in a packed bed at specific powers of 0.16 to 0.58 W/g. Dryout in the debris was achieved at powers from 0.42 to 0.58 W/g. Channels were induced in the bed and channeled bed dryout was achieved at powers of 1.06 to 1.77 W/g. Maximum temperatures in excess of 2500/sup 0/C were attained.

  10. Experimental method for reactor-noise measurements of effective beta. [LMFBR

    SciTech Connect

    Bennett, E.F.

    1981-09-01

    A variance-to-mean noise technique, modified to eliminate systematic errors from drifting of reactor power, has been used to infer integral values of effective beta for uranium and plutonium fueled fast reactor modk-ups. The measurement technique, including corrections for a finite detector-electrometer time response, is described together with preliminary beta measurement results.

  11. Preliminary considerations on developing IAEA technical safeguards for LMFBR power systems

    SciTech Connect

    Persiani, P. J.

    1980-09-01

    Nuclear fuel cycles safeguards should be considered in the dynamic context of a world deployment of various reactor types and varying availability of fuel-cycle services. There will be a close interaction between thermal-reactor cycles and the future deployment of fast breeders. The quantitites of plutonium and the reprocessing, conversion, fabrication, and storage methods of the fuel for the fast breeders will have a significant impact on safeguards techniques. The approach to the fast breeder fuel cycle safeguards follows the general safeguards system approach proposed by the IAEA. Objective of IAEA safeguards is the detection of diversion of nuclear material and deterrence of such diversion. To achieve independent verification of material balance accountancy requires the capability to monitor inventory status and verify material flows and quantities of all nuclear materials subject to safeguards. Containment and surveillance measures are applied to monitor key measurement points, maintain integrity of material balance, and complement material accountancy. The safeguards study attempts to develop a generic reference IAEA Safeguards System and explores various system options using containment/surveillance and material accountancy instrumentation and integrated systems designs.

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

    SciTech Connect

    De Volpi, A.

    1981-01-01

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

  13. Aerosol release and transport program. Quarterly progress report, October-December 1981. [LMFBR; PWR; BWR

    SciTech Connect

    Adams, R. E.; Tobias, M. L.

    1982-05-01

    This report summarizes progress for the Aerosol Release and Transport Program sponsored by the Nuclear Regulatory Commission's Office of Nuclear Regulatory Research, Division of Accident Evaluation, for the period October-December 1981. Topics discussed include (1) under-sodium tests in the Fuel Aerosol Simulant Test (FAST) Facility, (2) U/sub 3/O/sub 8/ and Fe/sub 2/O/sub 3/ in steam (light-water reactor accident) aerosol experiments in the Nuclear Safety Pilot Plant, (3) generation and characterization of cadmium and CdO aerosols in the basic aerosol experimental program, (4) core-melt tests of Zircaloy-clad fuel capsules, (5) initial results of a piston-model bubble oscillation code allowing liquid bypass, and (6) calculations with the UVABUBL code to compare with underwater and under-sodium period measurements in FAST experiments.

  14. Cadmium Depletion Impacts on Hardening Neutron6 Spectrum for Advanced Fuel Testing in ATR

    SciTech Connect

    Gray S. Chang

    2011-05-01

    For transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products effectively is in a fast neutron spectrum reactor. In the absence of a fast spectrum test reactor in the United States of America (USA), initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. A test region is achieved with a Cadmium (Cd) filter which can harden the neutron spectrum to a spectrum similar (although still somewhat softer) to that of the liquid metal fast breeder reactor (LMFBR). A fuel test loop with a Cd-filter has been installed within the East Flux Trap (EFT) of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). A detailed comparison analyses between the cadmium (Cd) filter hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum have been performed using MCWO. MCWO is a set of scripting tools that are used to couple the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2.2. The MCWO-calculated results indicate that the Cd-filter can effectively flatten the Rim-Effect and reduce the linear heat rate (LHGR) to meet the advanced fuel testing project requirements at the beginning of irradiation (BOI). However, the filtering characteristics of Cd as a strong absorber quickly depletes over time, and the Cd-filter must be replaced for every two typical operating cycles within the EFT of the ATR. The designed Cd-filter can effectively depress the LHGR in experimental fuels and harden the neutron spectrum enough to adequately flatten the Rim Effect in the test region.

  15. LMFBR aerosol release and transport program. Quarterly progress report, July-September 1981

    SciTech Connect

    Kress, T.S.; Tobias, M.L.

    1982-01-01

    This report summarizes progress for the Aerosol Release and Transport Program sponsored by the Office of Nuclear Regulatory Research, Division of Accident Evaluation of the Nuclear Regulatory Commission for the period July-September 1981. Topics discussed include (1) preparations for under-sodium tests at the Fast Aerosol Simulant Test Facility, (2) progress in interpretation of Oak Ridge National Laboratory-Sandia Laboratory normalization test results, (3) U/sub 3/O/sub 8/ in steam (light-water reactor accident) aerosol experiments conducted in the Nuclear Safety Power Plant, (4) experiments on B/sub 2/O/sub 3/ and SiO/sub 2/ aerosols at the Containment Research Installation-II Facility, (5) fuel-melting tests in small-scale experimental facilities for the core-melt aerosol program, (6) analytical comparison of simple adiabatic nonlinear and linear analytical models of bubble oscillation phenomena with experimental data.

  16. Solid oxide fuel cell with multi-unit construction and prismatic design

    SciTech Connect

    McPheeters, Charles C.; Dees, Dennis W.; Myles, Kevin M.

    1997-12-01

    A single cell unit of a solid oxide fuel cell is described that is individually fabricated and sintered prior to being connected to adjacent cells to form a solid oxide fuel cell . The single cell unit is comprised of a shaped anode sheet positioned between a flat anode sheet and an anode-electrolyte-cathode (A/E/C) sheet, and a shaped cathode sheet positioned between the A/E/C sheet and a cathode-interconnect-anode (C/I/A) sheet. An alternate embodiment comprises a shaped cathode sheet positioned between an A/E/C sheet and a C/I/A sheet. The shaped sheets form channels for conducting reactant gases. Each single cell unit is individually sintered to form a finished sub-assembly. The finished sub-assemblies are connected in electrical series by interposing connective material between the end surfaces of adjacent cells, whereby individual cells may be inspected for defects and interchanged with non-defective single cell units.

  17. Solid oxide fuel cell with multi-unit construction and prismatic design

    DOEpatents

    McPheeters, C.C.; Dees, D.W.; Myles, K.M.

    1999-03-16

    A single cell unit of a solid oxide fuel cell is described that is individually fabricated and sintered prior to being connected to adjacent cells to form a solid oxide fuel cell. The single cell unit is comprised of a shaped anode sheet positioned between a flat anode sheet and an anode-electrolyte-cathode (A/E/C) sheet, and a shaped cathode sheet positioned between the A/E/C sheet and a cathode-interconnect-anode (C/I/A) sheet. An alternate embodiment comprises a shaped cathode sheet positioned between an A/E/C sheet and a C/I/A sheet. The shaped sheets form channels for conducting reactant gases. Each single cell unit is individually sintered to form a finished sub-assembly. The finished sub-assemblies are connected in electrical series by interposing connective material between the end surfaces of adjacent cells, whereby individual cells may be inspected for defects and interchanged with non-defective single cell units. 7 figs.

  18. Solid oxide fuel cell with multi-unit construction and prismatic design

    DOEpatents

    McPheeters, Charles C.; Dees, Dennis W.; Myles, Kevin M.

    1999-01-01

    A single cell unit of a solid oxide fuel cell that is individually fabricated and sintered prior to being connected to adjacent cells to form a solid oxide fuel cell. The single cell unit is comprised of a shaped anode sheet positioned between a flat anode sheet and an anode-electrolyte-cathode (A/E/C) sheet, and a shaped cathode sheet positioned between the A/E/C sheet and a cathode-interconnect-anode (C/I/A) sheet. An alternate embodiment comprises a shaped cathode sheet positioned between an A/E/C sheet and a C/I/A sheet. The shaped sheets form channels for conducting reactant gases. Each single cell unit is individually sintered to form a finished sub-assembly. The finished sub-assemblies are connected in electrical series by interposing connective material between the end surfaces of adjacent cells, whereby individual cells may be inspected for defects and interchanged with non-defective single cell units.

  19. Fossil fuels -- future fuels

    SciTech Connect

    1998-03-01

    Fossil fuels -- coal, oil, and natural gas -- built America`s historic economic strength. Today, coal supplies more than 55% of the electricity, oil more than 97% of the transportation needs, and natural gas 24% of the primary energy used in the US. Even taking into account increased use of renewable fuels and vastly improved powerplant efficiencies, 90% of national energy needs will still be met by fossil fuels in 2020. If advanced technologies that boost efficiency and environmental performance can be successfully developed and deployed, the US can continue to depend upon its rich resources of fossil fuels.

  20. Opportunity fuels

    SciTech Connect

    Lutwen, R.C.

    1994-12-31

    Opportunity fuels - fuels that can be converted to other forms of energy at lower cost than standard fossil fuels - are discussed in outline form. The type and source of fuels, types of fuels, combustability, methods of combustion, refinery wastes, petroleum coke, garbage fuels, wood wastes, tires, and economics are discussed.

  1. Effect of yttrium additions on the elevated-temperature tensile properties and hardness of an advanced iron-nickel-chromium LMFBR cladding and duct alloy

    SciTech Connect

    Song, M.H.

    1981-10-01

    The effect of the addition of yttrium on the elevated temperature tensile properties and hardness of an Fe-34% Ni-12% Cr candidate LMFBR cladding and duct alloy was investigated. Tensile tests were performed from room temperature to 800/sup 0/C in 100/sup 0/C steps at strain rates of 2.2 x 10/sup -3/ and 2.2 x 10/sup -4/ sec/sup -1/. Hardness tests were performed from room temperature to 850/sup 0/C in 50/sup 0/C steps. The addition of 0.1% yttrium decreased the yield stress and ultimate tensile stress in the test temperature range employed. Hardness also decreased over this test temperature range. In tensile tests, dynamic strain aging behavior occurred both for the undoped and doped alloy in the temperature range from 200 to 600/sup 0/C and 300 to 600/sup 0/C for the lower and higher strain rate, respectively.

  2. On the Corrosion adequacy of the 2 1/4 CR-1Mo steel for LMFBR steam generation system service. Critical literature survey

    SciTech Connect

    Zima, G.E.

    1980-05-01

    The focus of this review is on the long-term serviceability of 2 1/4-1 Mo steel under the waterside environmental conditions presented in the steam generator of an LMFBR commercial scale plant. The basic question related to material behavior is to what extent the water side physico-chemical environment will affect the favorable performance of a given material under operating experience. In present light water reactors, the steam generator corrosion problems in part are attributable to complex interactions between the localized secondary environment and the mechanical design of the components (i.e., tube/tube support crevice, tube/tubesheet crevice, flow pattern, etc.) in the steam generating system.

  3. Results and code predictions for ABCOVE (aerosol behavior code validation and evaluation) aerosol code validation: Test AB6 with two aerosol species. [LMFBR

    SciTech Connect

    Hilliard, R K; McCormack, J C; Muhlestein, L D

    1984-12-01

    A program for aerosol behavior code validation and evaluation (ABCOVE) has been developed in accordance with the LMFBR Safety Program Plan. The ABCOVE program is a cooperative effort between the USDOE, the USNRC, and their contractor organizations currently involved in aerosol code development, testing or application. The second large-scale test in the ABCOVE program, AB6, was performed in the 850-m/sup 3/ CSTF vessel with a two-species test aerosol. The test conditions simulated the release of a fission product aerosol, NaI, in the presence of a sodium spray fire. Five organizations made pretest predictions of aerosol behavior using seven computer codes. Three of the codes (QUICKM, MAEROS and CONTAIN) were discrete, multiple species codes, while four (HAA-3, HAA-4, HAARM-3 and SOFIA) were log-normal codes which assume uniform coagglomeration of different aerosol species. Detailed test results are presented and compared with the code predictions for seven key aerosol behavior parameters.

  4. Evaluation and improvement on external-hazard proof of JSFR fuel handling system

    SciTech Connect

    Katoh, A.; Chikazawa, Y.; Uzawa, M.

    2012-07-01

    Responding to the the Fukushima Dai-ichi nuclear power plant (1F-NPP) accident, the earthquake and the tsunami proof of the fuel handling system (FHS) in Japan sodium-cooled fast reactor (JSFR) is studied. In the earthquake proof estimation, the margin of seismic resistance against the earthquake of the 1F-envelop condition and the sloshing behavior in the EVST is estimated. In terms of the tsunami proof, the scenario to lead fuel subassemblies into the stable cooling state and the potential of the cooling system is introduced in case of loss of the emergency power supply. As a result, it is clear that JSFR FHS originally could already be prepared to have the potential to prevent the release of radioactive material. (authors)

  5. Fuel pin

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

    1987-11-24

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  6. Inherent safety advantages of carbide fuel systems and technical issues regarding natural convection in LMRs

    SciTech Connect

    Barthold, W.P.

    1984-08-01

    The scope of work is to summarize inherent safety advantages that are unique to the use of a carbide based fuel system and to summarize the technical issues regarding natural convection flow in LMFBR cores. As discussed in this report, carbide fuel provides the designer with far greater flexibility than oxide fuel. Carbide fuel systems can be designed to eliminate major accident initiators. They turn quantitative advantages into a qualitative advantage. The author proposed to LANL a series of core design and component concepts that would greatly enhance the safety of carbide over oxide systems. This report cites a series of safety advantages which potentially exist for a carbide fuel system. Natural convection issues have not been given much attention in the past. Only during the last few years has this issue been addressed in some detail. Despite claims to the contrary by some of the LMR contractors, the author does not think that the natural convection phenomena is fully understood. Some of the approximations made in natural convection transient analyses have probably a greater impact on calculated transient temperatures than the effects under investigation. Only integral in-pile experimental data and single assembly out-of-pile detailed data are available for comparisons with analytical models and correlations. Especially for derated cores, the natural convection capability of a LMR should be far superior to that of a LWR. The author ranks the natural convection capability of the LMR as the most important inherent safety feature.

  7. Crossflow force transducer. [LMFBR

    SciTech Connect

    Mulcahy, T M

    1982-05-01

    A force transducer for measuring lift and drag coefficients for a circular cylinder in turbulent water flow is presented. In addition to describing the actual design and construction of the strain-gauged force- ring based transducer, requirements for obtained valid fluid force test data are discussed, and pertinent flow test experience is related.

  8. The Need for Confirmatory Experiments on the Radioactive Source Term from Potential Sabotage of Spent Nuclear Fuel Casks

    SciTech Connect

    PHILBIN, JEFFREY S.; HOOVER, MARK D.; NEWTON, GEORGE J.

    2002-04-01

    A technical review is presented of experiment activities and state of knowledge on air-borne, radiation source terms resulting from explosive sabotage attacks on spent reactor fuel subassemblies in shielded casks. Current assumptions about the behavior of irradiated fuel are largely based on a limited number of experimental results involving unirradiated, depleted uranium dioxide ''surrogate'' fuel. The behavior of irradiated nuclear fuel subjected to explosive conditions could be different from the behavior of the surrogate fuel, depending on the assumptions made by the evaluator. Available data indicate that these potential differences could result in errors, and possible orders-of-magnitude overestimates of aerosol dispersion and potential health effects from sabotage attacks. Furthermore, it is suggested that the current assumptions used in arriving at existing regulations for the transportation and storage of spent fuel in the U.S. are overly conservative. This, in turn, has led to potentially higher-than-needed operating expenses for those activities. A confirmatory experimental program is needed to develop a realistic correlation between source terms of irradiated fuel and unirradiated fuel. The motivations for performing the confirmatory experimental program are also presented.

  9. Opportunity fuels

    SciTech Connect

    Lutwen, R.C.

    1996-12-31

    The paper consists of viewgraphs from a conference presentation. A comparison is made of opportunity fuels, defined as fuels that can be converted to other forms of energy at lower cost than standard fossil fuels. Types of fuels for which some limited technical data is provided include petroleum coke, garbage, wood waste, and tires. Power plant economics and pollution concerns are listed for each fuel, and compared to coal and natural gas power plant costs. A detailed cost breakdown for different plant types is provided for use in base fuel pricing.

  10. Synthetic fuels

    SciTech Connect

    Sammons, V.O.

    1980-01-01

    This guide is designed for those who wish to learn more about the science and technology of synthetic fuels by reviewing materials in the collections of the Library of Congress. This is not a comprehensive bibliography, it is designed to put the reader on target. Subject headings used by the Library of Congress under which books on synthetic fuels can be located are: oil-shale industry; oil-shales; shale oils; synthetic fuels; synthetic fuels industry; coal gasification; coal liquefaction; fossil fuels; hydrogen as fuel; oil sands; petroleum, synthesis gas; biomass energy; pyrolysis; and thermal oil recovery. Basic texts, handbooks, government publications, journals, etc. were included. (DP)

  11. Synthetic Fuel

    ScienceCinema

    Idaho National Laboratory - Steve Herring, Jim O'Brien, Carl Stoots

    2010-01-08

    Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhouse gass Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhous

  12. Synthetic Fuel

    SciTech Connect

    Idaho National Laboratory - Steve Herring, Jim O'Brien, Carl Stoots

    2008-03-26

    Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhouse gass Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhous

  13. EBR-II transient operation and test capabilities

    SciTech Connect

    Seidel, B.R.; Cutforth, D.C.; Lentz, G.L.; Lambert, J.D.B.

    1983-01-01

    Experimental fuel pins intended for eventual use in LMFBR's have been irradiated for many years in fast test reactors. A wealth of data have been obtained on their performance under steady-state conditions, and fuel-pin performance codes have been developed to predict their behavior. In parallel, safety tests of fuel pins to explore behavior under accident conditions have been performed in transient reactors like TREAT in the US, and CABRI in France. These two types of testing generally have had different aims and have tended to produce results which are not reconcilable with a common modeling code, such as a LIFE or COMETHE, in the middle ground between normal and off-normal conditions. But as the licensing and commercialization of LMFBR's approaches, the attention and needs of the fuel-pin designer and licenser have focused on this middle ground between steady-state and accident testing of fuel pins and subassemblies. Preparations and now capability for operational reliability testing at EBR-II have been the subject of papers at recent conferences. This paper updates the status of those preparations to the present time when the ORT program is about to begin.

  14. Alternate fuels

    SciTech Connect

    Ryan, T.W.; Worthen, R.P.

    1981-02-01

    The escalating oil prices and shortages of petroleum based fuels for transportation have made research work on various fuel alternatives, especially for transportation engines, a priority of both the private and public sectors. This book contains 18 papers on this subject. The range of options from the development of completely non-petroleum-based fuels and engines to the use of various non-petroleum gasoline and diesel fuel extenders and improvers are discussed.

  15. Fossil Fuels.

    ERIC Educational Resources Information Center

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  16. Examination of frit vent from Sixty-Watt Heat Source simulant fueled clad vent set

    SciTech Connect

    Ulrich, G.B.

    1995-11-01

    The flow rate and the metallurgical condition of a frit vent from a simulant-fueled clad vent set (CVS) that had been hot isostatically pressed (HIP) for the Sixty-Watt Heat Source program were evaluated. The flow rate form the defueled vent cup subassembly was reduced approximately 25% from the original flow rate. No obstructions were found to account for the reduced flow rate. Measurements indicate that the frit vent powder thickness was reduced about 30%. Most likely, the powder was compressed during the HIP operation, which increased the density of the powder layer and thus reduced the flow rate of the assembly. All other observed manufacturing attributes appeared to be normal, but the vent hole activation technique needs further refinement before it is used in applications requiring maximum CVS integrity.

  17. Alternative fuels

    NASA Technical Reports Server (NTRS)

    Grobman, J. S.; Butze, H. F.; Friedman, R.; Antoine, A. C.; Reynolds, T. W.

    1977-01-01

    Potential problems related to the use of alternative aviation turbine fuels are discussed and both ongoing and required research into these fuels is described. This discussion is limited to aviation turbine fuels composed of liquid hydrocarbons. The advantages and disadvantages of the various solutions to the problems are summarized. The first solution is to continue to develop the necessary technology at the refinery to produce specification jet fuels regardless of the crude source. The second solution is to minimize energy consumption at the refinery and keep fuel costs down by relaxing specifications.

  18. Proposed power upgrade of the Hot Fuel Examination Facility's neutron radiography reactor. [NRAD reactor

    SciTech Connect

    Pruett, D.P.; Richards, W.J.; Heidel, C.C.

    1984-01-01

    The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both non-destructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the non-destructive examination techniques utilized at HFEF is neutron radiography. Neutron radiography is provided by the NRAD reactor facility, which is located beneath the HFEF hot cell. The NRAD reactor is a TRIGA reactor and is operated at a steady state power level of 250 kW solely for neutron radiography and the development of radiography techniques. When the NRAD facility was designed and constructed, an operating power level of 250 kW was considered to be adequate for obtaining radiographs of the type of specimens envisaged at that time. A typical radiograph required approximately a twenty-minute exposure time. Specimens were typically single fuel rods placed in an aluminum tray. Since that time, however, several things have occurred that have tended to increase radiography exposure times to as much as 90 minutes each. In order to decrease exposure times, the reactor power level is to be increased from 250 kw to 1 MW. This increase in power will necessitate several engineering and design changes. These changes are described.

  19. An evaluation of retention and disposal options for tritium in fuel reprocessing

    SciTech Connect

    Benjamin, R.W.; Hampson, D.C.

    1987-12-31

    This report assesses the possible options for retention of tritium and its ultimate disposal during future reprocessing of irradiated oxide fuels discharged from light water reactors (LWRs) and liquid metal fast breeder reactors (LMFBRs). The assessment includes an appraisal of the state of the retention and disposal options, an estimate of the dose commitments to the general public, an estimation of the incremental costs of the several retention and disposal options, and the potential reduction of the dose commitments resulting from retention and disposal of the tritium. The assessment is based upon an extensive study of tritium retention in reprocessing completed in 1982 by Grimes et al. Two plants were assumed, one to process LWR oxide fuel and the other to process LMFBR fuel. In each base case plant the tritium was vaporized to the atmosphere. Each of the hypothetical plants was assumed to be constructed during the 1990`s and to operate for a 20-year lifetime beginning in the year 2000 at a rate of 1,500 metric tons of heavy metal (MTHM) per 300-d year. In addition to the base case (Case 1), six other cases which included tritium retention options were examined. Although many of the features of the base-case plants remain unchanged in the tritium retention options, each case requires some additions, deletions, and modifications of portions of the plants. The retained tritium must also be managed and disposed of in a manner that is environmentally acceptable.

  20. Fuel cells 101

    SciTech Connect

    Hirschenhofer, J.H.

    1999-07-01

    This paper discusses the various types of fuel cells, the importance of cell voltage, fuel processing for natural gas, cell stacking, fuel cell plant description, advantages and disadvantages of the types of fuel cells, and applications. The types covered include: polymer electrolyte fuel cell, alkaline fuel cell, phosphoric acid fuel cell; molten carbonate fuel cell, and solid oxide fuel cell.

  1. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    NASA Astrophysics Data System (ADS)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  2. Analyzing Nuclear Fuel Cycles from Isotopic Ratios of Waste Products Applicable to Measurement by Accelerator Mass Spectrometry

    SciTech Connect

    Biegalski, S R; Whitney, S M; Buchholz, B

    2005-08-24

    An extensive study was conducted to determine isotopic ratios of nuclides in spent fuel that may be utilized to reveal historical characteristics of a nuclear reactor cycle. This forensic information is important to determine the origin of unknown nuclear waste. The distribution of isotopes in waste products provides information about a nuclear fuel cycle, even when the isotopes of uranium and plutonium are removed through chemical processing. Several different reactor cycles of the PWR, BWR, CANDU, and LMFBR were simulated for this work with the ORIGEN-ARP and ORIGEN 2.2 codes. The spent fuel nuclide concentrations of these reactors were analyzed to find the most informative isotopic ratios indicative of irradiation cycle length and reactor design. Special focus was given to long-lived and stable fission products that would be present many years after their creation. For such nuclides, mass spectrometry analysis methods often have better detection limits than classic gamma-ray spectroscopy. The isotopic ratios {sup 151}Sm/{sup 146}Sm, {sup 149}Sm/{sup 146}Sm, and {sup 244}Cm/{sup 246}Cm were found to be good indicators of fuel cycle length and are well suited for analysis by accelerator mass spectroscopy.

  3. Fuel injector

    DOEpatents

    Lambeth, Malcolm David Dick

    2001-02-27

    A fuel injector comprises first and second housing parts, the first housing part being located within a bore or recess formed in the second housing part, the housing parts defining therebetween an inlet chamber, a delivery chamber axially spaced from the inlet chamber, and a filtration flow path interconnecting the inlet and delivery chambers to remove particulate contaminants from the flow of fuel therebetween.

  4. Visualizing Coolant Flow in Sodium Reactor Subassemblies

    SciTech Connect

    2010-01-01

    Uniformity of temperature controls peak power output. Interchannel cross-flow is the principal cross-assembly energy transport mechanism. The areas of fastest flow all occur at the exterior of the assembly. Further, the fast moving region winds around the assembly in a continuous swath. This Nek5000 simulation uses an unstructured mesh with over one billion grid points, resulting in five billion degrees of freedom per time slice. High speed patches of turbulence due to vertex shedding downstream of the wires persist for about a quarter of the wire-wrap periodic length. Credits: Science: Paul Fisher and Aleks Obabko, Argonne National Laboratory
 Visualization: Hank Childs and Janet Jacobsen, Lawrence Berkeley National Laboratory

 This research used resources of the Argonne Leadership Computing Facility at Argonne National Laboratory, which is supported by the Office of Science of the U.S. Dept. of Energy under contract DE-AC02-06CH11357. This research was sponsored by the Department of Energy's Office of Nuclear Energy's NEAMS program.

  5. Wind turbine generator with improved operating subassemblies

    DOEpatents

    Cheney, Jr., Marvin C.

    1985-01-01

    A wind turbine includes a yaw spring return assembly to return the nacelle from a position to which it has been rotated by yawing forces, thus preventing excessive twisting of the power cables and control cables. It also includes negative coning restrainers to limit the bending of the flexible arms of the rotor towards the tower, and stop means on the rotor shaft to orient the blades in a vertical position during periods when the unit is upwind when the wind commences. A pendulum pitch control mechanism is improved by orienting the pivot axis for the pendulum arm at an angle to the longitudinal axis of its support arm, and excessive creep is of the synthetic resin flexible beam support for the blades is prevented by a restraining cable which limits the extent of pivoting of the pendulum during normal operation but which will permit further pivoting under abnormal conditions to cause the rotor to stall.

  6. Fuel cell-fuel cell hybrid system

    DOEpatents

    Geisbrecht, Rodney A.; Williams, Mark C.

    2003-09-23

    A device for converting chemical energy to electricity is provided, the device comprising a high temperature fuel cell with the ability for partially oxidizing and completely reforming fuel, and a low temperature fuel cell juxtaposed to said high temperature fuel cell so as to utilize remaining reformed fuel from the high temperature fuel cell. Also provided is a method for producing electricity comprising directing fuel to a first fuel cell, completely oxidizing a first portion of the fuel and partially oxidizing a second portion of the fuel, directing the second fuel portion to a second fuel cell, allowing the first fuel cell to utilize the first portion of the fuel to produce electricity; and allowing the second fuel cell to utilize the second portion of the fuel to produce electricity.

  7. Fuel ethanol

    SciTech Connect

    Not Available

    1989-02-01

    This report discusses the Omnibus Trade and Competitiveness Act of 1988 which requires GAO to examine fuel ethanol imports from Central America and the Caribbean and their impact on the U.S. fuel ethanol industry. Ethanol is the alcohol in beverages, such as beer, wine, and whiskey. It can also be used as a fuel by blending with gasoline. It can be made from renewable resources, such as corn, wheat, grapes, and sugarcane, through a process of fermentation. This report finds that, given current sugar and gasoline prices, it is not economically feasible for Caribbean ethanol producers to meet the current local feedstock requirement.

  8. 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)

  9. Fuel composition

    SciTech Connect

    Johnson, T.H.

    1990-06-26

    This patent describes a motor fuel composition. It comprises: a mixture of hydrocarbons in the gasoline boiling range containing a deposit preventing or reducing effective amount of poly(olefin)-N-substituted- carbamate.

  10. Unconventional fuel: Tire derived fuel

    SciTech Connect

    Hope, M.W.

    1995-09-01

    Material recovery of scrap tires for their fuel value has moved from a pioneering concept in the early 1980`s to a proven and continuous use in the United States` pulp and paper, utility, industrial, and cement industry. Pulp and paper`s use of tire derived fuel (TDF) is currently consuming tires at the rate of 35 million passenger tire equivalents (PTEs) per year. Twenty mills are known to be burning TDF on a continuous basis. The utility industry is currently consuming tires at the rate of 48 million PTEs per year. Thirteen utilities are known to be burning TDF on a continuous basis. The cement industry is currently consuming tires at the rate of 28 million PTEs per year. Twenty two cement plants are known to be burning TDF on a continuous basis. Other industrial boilers are currently consuming tires at the rate of 6.5 million PTEs per year. Four industrial boilers are known to be burning TDF on a continuous basis. In total, 59 facilities are currently burning over 117 million PTEs per year. Although 93% of these facilities were not engineered to burn TDF, it has become clear that TDF has found acceptance as a supplemental fuel when blending with conventional fuels in existing combustion devices designed for normal operating conditions. The issues of TDF as a supplemental fuel and its proper specifications are critical to the successful development of this fuel alternative. This paper will focus primarily on TDF`s use in a boiler type unit.

  11. Alternative jet aircraft fuels

    NASA Technical Reports Server (NTRS)

    Grobman, J.

    1979-01-01

    Potential changes in jet aircraft fuel specifications due to shifts in supply and quality of refinery feedstocks are discussed with emphasis on the effects these changes would have on the performance and durability of aircraft engines and fuel systems. Combustion characteristics, fuel thermal stability, and fuel pumpability at low temperature are among the factors considered. Combustor and fuel system technology needs for broad specification fuels are reviewed including prevention of fuel system fouling and fuel system technology for fuels with higher freezing points.

  12. 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)

  13. Fuel compositions

    SciTech Connect

    Zaweski, E.F.; Niebylski, L.M.

    1986-09-23

    This patent describes a distillate fuel for indirect injection compression ignition engines containing at least the combination of (i) organic nitrate ignition accelerator, and (ii) an additive selected from the group consisting of alkenyl substituted succinimide, alkenyl substituted succinamide and mixtures thereof. The alkenyl substituent contains about 12-36 carbon atoms, the additive being made by the process comprising (a) isomerizing the double bond of an ..cap alpha..-olefin containing about 12-36 carbon atoms to obtain a mixture of internal olefins, (b) reacting the mixture of internal olefins with maleic acid, anhydride or ester to obtain an intermediate alkenyl substituted succinic acid, anhydride or ester, and (c) reacting the intermediate with ammonia to form a succinimide, succinamide or mixture thereof. The combination is present in an amount sufficient to minimize the coking characteristics of such fuel, especially throttling nozzle coking in the prechambers or swirl chambers of indirect injection compression ignition engines operated on such fuel.

  14. Reforming of fuel inside fuel cell generator

    DOEpatents

    Grimble, R.E.

    1988-03-08

    Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream 1 and spent fuel stream 2. Spent fuel stream 1 is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream 1 and exhaust stream 2, and exhaust stream 1 is vented. Exhaust stream 2 is mixed with spent fuel stream 2 to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells. 1 fig.

  15. Reforming of fuel inside fuel cell generator

    DOEpatents

    Grimble, Ralph E.

    1988-01-01

    Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream I and spent fuel stream II. Spent fuel stream I is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream I and exhaust stream II, and exhaust stream I is vented. Exhaust stream II is mixed with spent fuel stream II to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells.

  16. 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)

  17. Fuels characterization studies. [jet fuels

    NASA Technical Reports Server (NTRS)

    Seng, G. T.; Antoine, A. C.; Flores, F. J.

    1980-01-01

    Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

  18. Alcohol fuels

    SciTech Connect

    Not Available

    1981-07-01

    The API publication 4312 reports a detailed study carried out by Battelle on the energy balances for five alcohol-fuel-producing technologies. The results indicate that processes for producing ethanol from corn are net consumers of energy while ethanol from sugar cane and methanol from wood are net energy producers.

  19. Nuclear Fuels.

    ERIC Educational Resources Information Center

    Nash, J. Thomas

    1983-01-01

    Trends in and factors related to the nuclear industry and nuclear fuel production are discussed. Topics addressed include nuclear reactors, survival of the U.S. uranium industry, production costs, budget cuts by the Department of Energy and U.S. Geological survey for resource studies, mining, and research/development activities. (JN)

  20. Fuel Cells

    ERIC Educational Resources Information Center

    Hawkins, M. D.

    1973-01-01

    Discusses the theories, construction, operation, types, and advantages of fuel cells developed by the American space programs. Indicates that the cell is an ideal small-scale power source characterized by its compactness, high efficiency, reliability, and freedom from polluting fumes. (CC)

  1. Future Fuel.

    ERIC Educational Resources Information Center

    Stover, Del

    1991-01-01

    Tough new environmental laws, coupled with fluctuating oil prices, are likely to prompt hundreds of school systems to examine alternative fuels. Literature reviews and interviews with 45 government, education, and industry officials provided data for a comparative analysis of gasoline, diesel, natural gas, methanol, and propane. (MLF)

  2. Fuel issues for fuel cell vehicles

    SciTech Connect

    Borroni-Bird, C.E.

    1995-12-31

    In the near-term, infrastructure and energy density concerns dictate that the most appropriate fuel for a light-duty fuel cell vehicle is probably not hydrogen; there are also several concerns with using methanol, the generally accepted most convenient fuel. In order to accelerate fuel cell commercialization it may be necessary to use petroleum-based fuels and on-board fuel processors. In the near-term, this approach may reduce fuel cell system efficiency to a level comparable with advanced diesel engines but in the long-term fuel cells powered by hydrogen should be the most efficient and cleanest of all automotive powertrains.

  3. Fuel cells: A handbook

    NASA Astrophysics Data System (ADS)

    Kinoshita, K.; McLarnon, F. R.; Cairns, E. J.

    1988-05-01

    The purpose of this handbook is to present information describing fuel cells that is helpful to scientists, engineers, and technical managers who are not experienced in this technology, as well as to provide an update on the current technical status of the various types of fuel cells. Following the introduction, contents of this handbook are: fuel cell performance variables; phosphoric acid fuel cell; molten carbonate fuel cell; solid oxide fuel cell; alternative fuel cell technologies; fuel cell systems; and concluding remarks.

  4. Supplemental fuel vapor system

    SciTech Connect

    Foster, P.M.

    1991-01-08

    This patent describes a supplemental fuel system utilizing fuel vapor. It comprises: an internal combustion engine including a carburetor and an intake manifold; a fuel tank provided with air vents; a fuel conduit having a first end connected to the fuel tank and in communication with liquid fuel in the tank and a second end connected to the carburetor; the fuel conduit delivering the liquid fuel to the carburetor from the fuel tank; a fuel vapor conduit having a first end connected to the fuel tank at a location displaced from contact with the liquid fuel and a second end connected to a carbon canister; a PCV conduit having a first end connected to a pollution control valve and a second end connected to the intake manifold; and, an intermediate fuel vapor conduit having a first end connected to the fuel vapor conduit and a second end connected to the PCV conduit; wherein the air vents continuously provide air to the tank to mix with the liquid fuel and form fuel vapor. The fuel vapor drawn from the fuel tank by vacuum developed in the intake manifold and flows through the fuel vapor conduit. The intermediate fuel vapor conduit and the intake manifold to combustion chambers of the internal combustion engine so as to supplement fuel delivered to the engine by the fuel conduit. The liquid fuel and the fuel vapor constantly delivered to the engine during normal operation.

  5. ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

    SciTech Connect

    Lell, R. M.; Morman, J. A.; Schaefer, R.W.; McKnight, R.D.; Nuclear Engineering Division

    2009-02-23

    ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium

  6. Fuel composition

    SciTech Connect

    Johnson, T.H.

    1990-08-07

    This patent describes a concentrate suitable for use in liquid fuels in the gasoline boiling range. It comprises: from about 25 to about 500 ppm by weight of at least one poly(olefin)-N-substituted-carbamate; from about 0 to about 20 ppm by weight of a dehazer; and balance of diluent, boiling in the range from about 50{degrees}C. to about 232{degrees}C.

  7. System modelling to support accelerated fuel transfer rate at EBR-II

    SciTech Connect

    Imel, G.R.; Houshyar, A.; Planchon, H.P.; Cutforth, D.C.

    1995-06-01

    The Experimental Breeder Reactor-II (EBR-II) ia a 62.5 MW(th) liquid metal reactor operated by Argonne National Laboratory for The United States Department of Energy. The reactor is located near Idaho Falls, Idaho at the Argonne-West site (ANL-W). Full power operation was achieved in 1964,- the reactor operated continuously since that time until October 1994 in a variety of configurations depending on the programmatic mission. A three year program was initiated in October, 1993 to replace the 330 depleted uranium blanket subassemblies (S/As) with stainless steel reflectors. It was intended to operate the reactor during the three year blanket unloading program, followed by about a half year of driver fuel unloading. However, in the summer of 1994, Congress dictacted that EBR-II be shut down October 1, and complete defueling without operation. To assist in the planning for resources needed for this defueling campaign, a mathematical model of the fuel handling sequence was developed utilizing the appropriate reliability factors and inherent mm constraints of each stage of the process. The model allows predictions of transfer rates under different scenarios. Additionally, it has facilitated planning of maintenance activities, as well as optimization of resources regarding manpower and modification effort. The model and its application is described in this paper.

  8. Alcohol fuels

    SciTech Connect

    Not Available

    1990-07-01

    Ethanol is an alcohol made from grain that can be blended with gasoline to extend petroleum supplies and to increase gasoline octane levels. Congressional proposals to encourage greater use of alternative fuels could increase the demand for ethanol. This report evaluates the growth potential of the ethanol industry to meet future demand increases and the impacts increased production would have on American agriculture and the federal budget. It is found that ethanol production could double or triple in the next eight years, and that American farmers could provide the corn for this production increase. While corn growers would benefit, other agricultural segments would not; soybean producers, for example could suffer for increased corn oil production (an ethanol byproduct) and cattle ranchers would be faced with higher feed costs because of higher corn prices. Poultry farmers might benefit from lower priced feed. Overall, net farm cash income should increase, and consumers would see slightly higher food prices. Federal budget impacts would include a reduction in federal farm program outlays by an annual average of between $930 million (for double current production of ethanol) to $1.421 billion (for triple production) during the eight-year growth period. However, due to an partial tax exemption for ethanol blended fuels, federal fuel tax revenues could decrease by between $442 million and $813 million.

  9. Reactor vessel support system. [LMFBR

    DOEpatents

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  10. Fuel densifier converts biomass into fuel cubes

    SciTech Connect

    Not Available

    1982-02-01

    A new cost-effective means to produce clean-burning and low cost commercial and industrial fuel is being introduced by Columbia Fuel Densification Corp., Phoenix. The Columbia Commercial Hydraulic Fuel Densifier converts raw biomass materials such as wood chips, paper, peat moss and rice hulls into densified fuel cubes. The densifier is mobile and its operation is briefly outlined.

  11. 146. FUEL LINE TO SKID 2 (FUEL LOADER) IN FUEL ...

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

    146. FUEL LINE TO SKID 2 (FUEL LOADER) IN FUEL CONTROL ROOM (215), LSB (BLDG. 751). LIQUID NITROGEN/HELIUM HEAT EXCHANGER ON RIGHT. - Vandenberg Air Force Base, Space Launch Complex 3, Launch Pad 3 East, Napa & Alden Roads, Lompoc, Santa Barbara County, CA

  12. Aviation fuels outlook

    NASA Technical Reports Server (NTRS)

    Momenthy, A. M.

    1980-01-01

    Options for satisfying the future demand for commercial jet fuels are analyzed. It is concluded that the most effective means to this end are to attract more refiners to the jet fuel market and encourage development of processes to convert oil shale and coal to transportation fuels. Furthermore, changing the U.S. refineries fuel specification would not significantly alter jet fuel availability.

  13. Design study and comparative evaluation of JSFR failed fuel detection system

    SciTech Connect

    Aizawa, K.; Chikazawa, Y.; Ishikawa, N.; Kubo, S.; Okazaki, H.; Mito, M.; Tozawa, K.; Hayashi, M.

    2012-07-01

    A conceptual design study of an advanced sodium-cooled fast reactor JSFR has progressed in the 'Fast Reactor Cycle Technology Development (FaCT) 'project in Japan. JSFR has two failed fuel detection systems in the core. One is a failed fuel detection (FFD) system which continuously monitors a fission product from failed fuel subassembly. The other is a failed fuel detection and location (FFDL) system which locates when it receives signals from FFD. The FFD system consists of a FFD-DN which detects delayed neutron (DN) in sodium and a FFD-CG which detects fission products in the cover gas of the reactor vessel. In this study, requirements to the FFD-DN and the FFD-DN design to meet the requirements were investigated for the commercial and demonstration JSFR. In the commercial JSFR, a sampling type FFD which collects sodium from the reactor vessel by sampling lines for DN detectors was adopted. The performances have been investigated and confirmed by a fluid analysis in the reactor upper plenum. In the demonstration JSFR, the performance of DN detectors installed on the primary cold-leg piping has been confirmed. For the FFDL systems, experiences in the previous fast reactors and the R and D of FFDL system for JSFR were investigated. This study focuses on the Selector-Valve and the Tagging-Gas FFDL systems. Operation experiences of the Selector-valve FFDL system were accumulated in PFR and Phenix. Tagging-gas system experiences were accumulated in EBR-II and FFTF. The feasibility of both FFDL systems for JSFR was evaluated. (authors)

  14. Carburetor fuel discharge assembly

    SciTech Connect

    Yost, R.M.

    1993-06-29

    An improved carburetor for use on an internal combustion engine is described, the carburetor having an airflow passage and fuel discharge means for admitting fuel into the airflow passage for mixing the fuel with air flowing in the airflow passage to form a fuel/air mixture to be supplied to the combustion chamber(s) of the engine, the fuel discharge means including a fuel discharge assembly which comprises a hollow discharge tube and fuel supplying means connected to the discharge tube for admitting fuel into the interior of the discharge tube, wherein the discharge tube has a longitudinal internal bore in fluid communication with the fuel supplying means, wherein the internal bore extends between an inlet that is closest to the fuel supplying means and an outlet that is furthest from the fuel supplying means with the outlet of the bore being located within the airflow passage of the carburetor to supply fuel into this passage after the fuel passes from the fuel supplying means through the internal bore of the discharge tube, wherein the improvement relates to the fuel discharge assembly and comprises: a hollow fuel flow guide tube telescopically received inside the internal bore of the discharge tube, wherein the fuel flow guide tube extends from approximately the location of the inlet of the bore up at least a portion of the length of the bore towards the outlet of the bore to conduct fuel from the fuel supplying means into the bore of the discharge tube.

  15. Fuel processors for fuel cell APU applications

    NASA Astrophysics Data System (ADS)

    Aicher, T.; Lenz, B.; Gschnell, F.; Groos, U.; Federici, F.; Caprile, L.; Parodi, L.

    The conversion of liquid hydrocarbons to a hydrogen rich product gas is a central process step in fuel processors for auxiliary power units (APUs) for vehicles of all kinds. The selection of the reforming process depends on the fuel and the type of the fuel cell. For vehicle power trains, liquid hydrocarbons like gasoline, kerosene, and diesel are utilized and, therefore, they will also be the fuel for the respective APU systems. The fuel cells commonly envisioned for mobile APU applications are molten carbonate fuel cells (MCFC), solid oxide fuel cells (SOFC), and proton exchange membrane fuel cells (PEMFC). Since high-temperature fuel cells, e.g. MCFCs or SOFCs, can be supplied with a feed gas that contains carbon monoxide (CO) their fuel processor does not require reactors for CO reduction and removal. For PEMFCs on the other hand, CO concentrations in the feed gas must not exceed 50 ppm, better 20 ppm, which requires additional reactors downstream of the reforming reactor. This paper gives an overview of the current state of the fuel processor development for APU applications and APU system developments. Furthermore, it will present the latest developments at Fraunhofer ISE regarding fuel processors for high-temperature fuel cell APU systems on board of ships and aircrafts.

  16. Fuel cells: A survey

    NASA Technical Reports Server (NTRS)

    Crowe, B. J.

    1973-01-01

    A survey of fuel cell technology and applications is presented. The operating principles, performance capabilities, and limitations of fuel cells are discussed. Diagrams of fuel cell construction and operating characteristics are provided. Photographs of typical installations are included.

  17. Fuel economy of hydrogen fuel cell vehicles

    NASA Astrophysics Data System (ADS)

    Ahluwalia, Rajesh K.; Wang, X.; Rousseau, A.; Kumar, R.

    On the basis of on-road energy consumption, fuel economy (FE) of hydrogen fuel cell light-duty vehicles is projected to be 2.5-2.7 times the fuel economy of the conventional gasoline internal combustion engine vehicles (ICEV) on the same platforms. Even with a less efficient but higher power density 0.6 V per cell than the base case 0.7 V per cell at the rated power point, the hydrogen fuel cell vehicles are projected to offer essentially the same fuel economy multiplier. The key to obtaining high fuel economy as measured on standardized urban and highway drive schedules lies in maintaining high efficiency of the fuel cell (FC) system at low loads. To achieve this, besides a high performance fuel cell stack, low parasitic losses in the air management system (i.e., turndown and part load efficiencies of the compressor-expander module) are critical.

  18. Effect of hydrocarbon fuel type on fuel

    NASA Technical Reports Server (NTRS)

    Wong, E. L.; Bittker, D. A.

    1982-01-01

    A modified jet fuel thermal oxidation tester (JFTOT) procedure was used to evaluate deposit and sediment formation for four pure hydrocarbon fuels over the temperature range 150 to 450 C in 316-stainless-steel heater tubes. Fuel types were a normal alkane, an alkene, a naphthene, and an aromatic. Each fuel exhibited certain distinctive deposit and sediment formation characteristics. The effect of aluminum and 316-stainless-steel heater tube surfaces on deposit formation for the fuel n-decane over the same temperature range was investigated. Results showed that an aluminum surface had lower deposit formation rates at all temperatures investigated. By using a modified JFTOT procedure the thermal stability of four pure hydrocarbon fuels and two practical fuels (Jet A and home heating oil no. 2) was rated on the basis of their breakpoint temperatures. Results indicate that this method could be used to rate thermal stability for a series of fuels.

  19. Fuel processor for fuel cell power system

    DOEpatents

    Vanderborgh, Nicholas E.; Springer, Thomas E.; Huff, James R.

    1987-01-01

    A catalytic organic fuel processing apparatus, which can be used in a fuel cell power system, contains within a housing a catalyst chamber, a variable speed fan, and a combustion chamber. Vaporized organic fuel is circulated by the fan past the combustion chamber with which it is in indirect heat exchange relationship. The heated vaporized organic fuel enters a catalyst bed where it is converted into a desired product such as hydrogen needed to power the fuel cell. During periods of high demand, air is injected upstream of the combustion chamber and organic fuel injection means to burn with some of the organic fuel on the outside of the combustion chamber, and thus be in direct heat exchange relation with the organic fuel going into the catalyst bed.

  20. Fuel Processors for PEM Fuel Cells

    SciTech Connect

    Levi T. Thompson

    2008-08-08

    Fuel cells are being developed to power cleaner, more fuel efficient automobiles. The fuel cell technology favored by many automobile manufacturers is PEM fuel cells operating with H2 from liquid fuels like gasoline and diesel. A key challenge to the commercialization of PEM fuel cell based powertrains is the lack of sufficiently small and inexpensive fuel processors. Improving the performance and cost of the fuel processor will require the development of better performing catalysts, new reactor designs and better integration of the various fuel processing components. These components and systems could also find use in natural gas fuel processing for stationary, distributed generation applications. Prototype fuel processors were produced, and evaluated against the Department of Energy technical targets. Significant advances were made by integrating low-cost microreactor systems, high activity catalysts, π-complexation adsorbents, and high efficiency microcombustor/microvaporizers developed at the University of Michigan. The microreactor system allowed (1) more efficient thermal coupling of the fuel processor operations thereby minimizing heat exchanger requirements, (2) improved catalyst performance due to optimal reactor temperature profiles and increased heat and mass transport rates, and (3) better cold-start and transient responses.

  1. 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)

  2. Internal reforming fuel cell assembly with simplified fuel feed

    DOEpatents

    Farooque, Mohammad; Novacco, Lawrence J.; Allen, Jeffrey P.

    2001-01-01

    A fuel cell assembly in which fuel cells adapted to internally reform fuel and fuel reformers for reforming fuel are arranged in a fuel cell stack. The fuel inlet ports of the fuel cells and the fuel inlet ports and reformed fuel outlet ports of the fuel reformers are arranged on one face of the fuel cell stack. A manifold sealing encloses this face of the stack and a reformer fuel delivery system is arranged entirely within the region between the manifold and the one face of the stack. The fuel reformer has a foil wrapping and a cover member forming with the foil wrapping an enclosed structure.

  3. Fuel control system

    SciTech Connect

    Staniak, W.A.; Samuelson, R.E.; Moncelle, M.E.

    1986-10-14

    A fuel control system is described comprising: a fuel rack movable in opposite fuel-increasing and fuel-decreasing directions; a rack control member movable in opposite fuel-increasing and fuel-decreasing directions; servo system means for moving the fuel rack in response to movement of the rack control member an electrically energizable member movable in opposite fuel-increasing and fuel-decreasing directions, the electrically energizable member being urged to move in its fuel-decreasing direction when energized; first coupling means for connecting the electrically energizable member to the rack control member to move the rack control member in its fuel-decreasing direction in response to movement of the electrically energizable member in its fuel-decreasing direction; a mechanical governor control having a member movable in opposite fuel-increasing and fuel-decreasing directions; second coupling means for connecting the mechanical governor to the rack control member to move the rack control member in its fuel-decreasing direction in response to movement of the mechanical governor member in its fuel-decreasing direction; bias means for biasing the rack control member to move in its fuel-increasing direction.

  4. Fuel dissipater for pressurized fuel cell generators

    DOEpatents

    Basel, Richard A.; King, John E.

    2003-11-04

    An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a pressurized fuel cell generator (10) when the electrical power output of the fuel cell generator is terminated during transient operation, such as a shutdown; where, two electrically resistive elements (two of 28, 53, 54, 55) at least one of which is connected in parallel, in association with contactors (26, 57, 58, 59), a multi-point settable sensor relay (23) and a circuit breaker (24), are automatically connected across the fuel cell generator terminals (21, 22) at two or more contact points, in order to draw current, thereby depleting the fuel inventory in the generator.

  5. Bulk Fuel Man.

    ERIC Educational Resources Information Center

    Marine Corps Inst., Washington, DC.

    This student guide, one of a series of correspondence training courses designed to improve the job performance of members of the Marine Corps, deals with the skills needed by bulk fuel workers. Addressed in the four individual units of the course are the following topics: bulk fuel equipment, bulk fuel systems, procedures for handling fuels, and…

  6. FUEL ROD CLUSTERS

    DOEpatents

    Schultz, A.B.

    1959-08-01

    A cluster of nuclear fuel rods and a tubular casing therefor through which a coolant flows in heat-exchange contact with the fuel rods is described. The fuel rcds are held in the casing by virtue of the compressive force exerted between longitudinal ribs of the fuel rcds and internal ribs of the casing or the internal surfaces thereof.

  7. Microemulsion fuel system

    SciTech Connect

    Hazbun, E.A.; Schon, S.G.; Grey, R.A.

    1988-05-17

    A microemulsion fuel composition is described comprising: (a) a jet fuel, fuel oil or diesel hydrocarbon fuel; (b) about 3.0 to about 40% by weight water and/or methanol; and (c) a surface active amount of a combination of surface active agents consisting of: (1) tertiary butyl alcohol; and (2) at least one amphoteric; anionic, cationic or nonionic surfactant.

  8. Analysis of ICPP fuel storage rack inner tie and corner tie substructures

    SciTech Connect

    Nitzel, M.E.; Rahl, R.G.

    1996-01-01

    Finite element models were developed and analyses performed for the tie plate, inner tie block assembly, and corner tie block assembly of a 25 port fuel rack assembly designed for installation in Pool 1 of Building 666 at the Idaho Chemical Processing Plant. These models were specifically developed to investigate the adequacy of certain welds joining components of the fuel storage rack assembly. The work scope for the task was limited to an investigation of the stress levels in the subject subassemblies when subjected to seismic loads. Structural acceptance criteria used for the elastic calculations performed were as found in the overall rack design report as issued by the rack`s designer, Holtec International. Structural acceptance criteria used for the plastic calculations performed as part of this effort were as defined in Subsection NF and Appendix F of the ASME Boiler & Pressure Vessel Code. The results of the analyses will also apply to the 30 port fuel storage rack design that is also scheduled for installation in Pool 1 of ICPP 666. The results obtained from the analyses performed for this task indicate that the welds joining the inner tie block and corner tie block to the surrounding rack structure meet the acceptance criteria. Further, the structural members (plates and blocks) were also found to be within the allowable stress limits established by the acceptance criteria. The separate analysis performed on the inner tie plate confirmed the structural adequacy for both the inner tie plate, corner tie plate, and tie block bolts. The analysis results verified that the inner tie and corner tie block should be capable of transferring the expected seismic load without structural failure.

  9. Fuel cells seminar

    SciTech Connect

    1996-12-01

    This year`s meeting highlights the fact that fuel cells for both stationary and transportation applications have reached the dawn of commercialization. Sales of stationary fuel cells have grown steadily over the past 2 years. Phosphoric acid fuel cell buses have been demonstrated in urban areas. Proton-exchange membrane fuel cells are on the verge of revolutionizing the transportation industry. These activities and many more are discussed during this seminar, which provides a forum for people from the international fuel cell community engaged in a wide spectrum of fuel cell activities. Discussions addressing R&D of fuel cell technologies, manufacturing and marketing of fuel cells, and experiences of fuel cell users took place through oral and poster presentations. For the first time, the seminar included commercial exhibits, further evidence that commercial fuel cell technology has arrived. A total of 205 papers is included in this volume.

  10. Fuel transfer system

    DOEpatents

    Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool.