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Sample records for experimental fuel assemblies

  1. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    SciTech Connect

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  2. Fuel nozzle assembly

    DOEpatents

    Johnson, Thomas Edward; Ziminsky, Willy Steve; Lacey, Benjamin Paul; York, William David; Stevenson, Christian Xavier

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  3. Transfer of fuel assemblies

    SciTech Connect

    Vuckovich, M.; Burkett, J. P.; Sallustio, J.

    1984-12-11

    Fuel assemblies of a nuclear reactor are transferred during fueling or refueling or the like by a crane. The work-engaging fixture of the crane picks up an assembly, removes it from this slot, transfers it to the deposit site and deposits it in its slot at the deposit site. The control for the crane includes a strain gauge connected to the crane line which raises and lowers the load. The strain gauge senses the load on the crane. The signal from the strain gauge is compared with setpoints; a high-level setpoint, a low-level setpoint and a slack-line setpoint. If the strain gauge signal exceeds the high-level setpoint, the line drive is disabled. This event may occur during raising of a fuel assembly which encounters resistance. The high-level setpoint may be overridden under proper precautions. The line drive is also disabled if the strain gauge signal is less than the low-level setpoint. This event occurs when a fuel assembly being deposited contacts the bottom of its slot or an obstruction in, or at the entry to the slot. To preclude lateral movement and possible damage to a fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge exceets the lov-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than the slack-line set-point. This occurs when the lines has been set in slack-line setting. When the line is tensioned after slack-li ne setting, the traverse drive remains enabled only if the line has been disconnected from the fuel assembly.

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

    NASA Astrophysics Data System (ADS)

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

    2017-01-01

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

  5. Non-intrusive Experimental Study on Nuclear Fuel Assembly Response to Seismic Loads

    NASA Astrophysics Data System (ADS)

    Weichselbaum, Noah A.

    Experimental measurements of nuclear fuel bundle response to seismic loads have primarily been focused on the response of the structure. Forcing methods have included use of shake tables, however, the majority of work has used hydraulic actuators rigidly connected to a single spacer grid to force the fuel bundle. Structural measurements utilize such instruments as linear variable displacement transducers (LVDT) that are mounted on the structure. From these measurements it has been shown that fuel bundles in prototypical conditions, with an axial flow of 6 m/s, behave markedly different from fuel bundles in still water when there is external forcing on the core from an earthquake. It has also been shown that the structure and fluid are fully coupled. Thus more recently attention has been focused on fluid measurements in the bypass region around fuel bundles with external forcing with laser doppler velocimetry (LDV), which is a point wise fluid velocity measurement technique. This work describes a unique facility that has garnered a large experimental database of fully coupled fluid and structure measurements with time resolved particle image velocimetry (PIV) and digital image correlation (DIC) within a full height 6x6 fuel bundle exposed to seismic forcing on a large 6 degree of freedom shake table. A refractive index matched (RIM) vertical liquid tunnel is mounted on the shake table and houses the fuel bundle which is based on the geometry of a prototypical fuel bundle in a pressurized water reactor (PWR). PIV is obtained with high spatial resolution by rigidly mounting all optical equipment to the test section on the shake table, where the laser light is delivered through high power multi-mode step index fiber optics from a high powered Nd:YLF laser located 10 meters away from the test section. High temporal resolution for the PIV measurements is obtained with state of the art high speed CMOS cameras that record straight to hard drive allowing for increased

  6. Fuel cell sub-assembly

    DOEpatents

    Chi, Chang V.

    1983-01-01

    A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

  7. Fuel assembly for nuclear reactors

    DOEpatents

    Creagan, Robert J.; Frisch, Erling

    1977-01-01

    A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

  8. Experimental validation of CASMO-4E and CASMO-5M for radial fission rate distributions in a westinghouse SVEA-96 Optima2 BWR fuel assembly

    SciTech Connect

    Grimm, P.; Perret, G.

    2012-07-01

    Measured and calculated radial total fission rate distributions are compared for the three axial sections of a Westinghouse SVEA-96 Optima2 BWR fuel assembly, comprising 96, 92 and 84 fuel rods, respectively. The measurements were performed on a full-size fuel assembly in the PROTEUS zero-power experimental facility. The measured fission rates are compared to the results of the CASMO-4E and CASMO-5M fuel assembly codes. Detailed measured geometrical data were used in the models, and effects of the surrounding zones of the reactor were taken into account by correction factors derived from MCNPX calculations. The results of the calculations agree well with those of the experiments, with root-mean-square deviations between 1.2% and 1.5% and maximum deviations of 3-4%. The quality of the predictions by CASMO-4E and CASMO-5M is comparable. (authors)

  9. Experimental studies of local coolant hydrodynamics using a scaled model of cassette-type fuel assembly of a KLT-40S reactor

    NASA Astrophysics Data System (ADS)

    Dmitriev, S. M.; Barinov, A. A.; Varentsov, A. V.; Doronkov, D. V.; Solntsev, D. N.; Khrobostov, A. E.

    2016-08-01

    The results of experimental studies of local hydrodynamic and mass exchange characteristics of the coolant flow behind the spacer grid in the fuel assembly of a KLT-40S reactor are presented. The experiments were aimed at the investigation of representative domains of the fuel assembly with three tracer injection regions. The studies were performed at the aerodynamic test facility using the tracer gas diffusion method. According to the theory of hydrodynamic similarity, the obtained experimental results can be transferred to full-scale coolant flow conditions in standard fuel assemblies. The analysis of the tracer concentration propagation made it possible to determine in detail the flow pattern and find the main regularities and specific features of the coolant flow behind the plate spacer grid of KLT-40S fuel assembly. The hydraulic resistance coefficient of the spacer grid was experimentally determined. The coefficients of mass exchange between cells for representative cells of the displacer region in the KLT-40S fuel assembly were calculated for the first time; these results are presented in the form of the "mixing matrix." The results of studies of local coolant flow hydrodynamics in the KLT-40S fuel assembly are used at AO Afrikantov OKBM for estimation of thermotechnical reliability of active cores for reactors of floating nuclear power stations. The experimental data on hydrodynamic and mass exchange characteristics are included in the database for verification of CDF codes and detailed cell-wise calculation of the active core for KLT-40S reactor installation. The results of these studies can be used at FSUE RFNC-VNIIEF for testing and verification of domestic three-dimensional hydrodynamic CFD codes ("Logos") that are applied for substantiation of newly designed reactor installations. Practical recommendations on the application of the obtained results in thermohydraulic calculations of the active core for the KLT-40S reactor will be worked out. Proposals

  10. Fuel cell design and assembly

    DOEpatents

    Myerhoff, Alfred

    1984-01-01

    The present invention is directed to a novel bipolar cooling plate, fuel cell design and method of assembly of fuel cells. The bipolar cooling plate used in the fuel cell design and method of assembly has discrete opposite edge and means carried by the plate defining a plurality of channels extending along the surface of the plate toward the opposite edges. At least one edge of the channels terminates short of the edge of the plate defining a recess for receiving a fastener.

  11. Nuclear core and fuel assemblies

    DOEpatents

    Downs, Robert E.

    1981-01-01

    A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

  12. FUEL ROD ASSEMBLY

    DOEpatents

    Hutter, E.

    1959-09-01

    A cluster of nuclear fuel rods aod a tubular casing through which a coolant flows in heat-change contact with the ruel rods are described. The casting is of trefoil section and carries the fuel rods, each of which has two fin engaging the serrated fins of the other two fuel rods, whereby the fuel rods are held in the casing and are interlocked against relative longitudinal movement.

  13. Nuclear reactor composite fuel assembly

    DOEpatents

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  14. FUEL ASSEMBLY SHAKER TEST SIMULATION

    SciTech Connect

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

    2013-05-30

    This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through

  15. Reactivity measurements on an experimental assembly of 4. 31 wt % sup 235 U enriched UO sub 2 fuel rods arranged in a shipping cask geometry

    SciTech Connect

    Bierman, S.R.

    1989-10-01

    A research program was initiated for the US Department of Energy (DOE) Sandia National Laboratory Transportation Systems Development Department in 1982 to provide benchmark type experimental criticality data in support of the design and safe operations of nuclear fuel transportation systems. The overall objective of the program is to identify and provide the experimental data needed to form a consistent, firm, and complete data base for verifying calculational models used in the criticality analyses of nuclear transport and related systems. A report, PNL-6205, issued in June 1988 (Bierman 1988) covered measurement results obtained from a series of experimental assemblies (TIC-1, 2, 3 and 4) involving neutron flux traps. The results obtained on a fifth experimental assembly (TIC-5), modeled after a calculational problem of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI) Working Group, are covered in this report. 10 refs., 10 figs., 7 tabs.

  16. Nuclear fuel assembly wear sleeve

    SciTech Connect

    Cadwell, D.J.; Kmonk, S.

    1983-03-08

    An improved control rod guide tube for use in a fuel assembly in a nuclear reactor. The guide tube extends the complete length of the fuel assembly and has its upper end fastened in a cylindrical housing by swaging the guide tube material into grooves formed in the housing walls. To eliminate wear on the guide tube inner walls caused by hydraulic induced vibratory forces on a control rod adapted to move therein, a thin-walled chrome plated sleeve is threaded into the top end of the guide thimble and extends downwardly a distance sufficient to be engaged by the control rod during reactor operation. The sleeve serves as a highly resistant wear surface between the control rod and walls on the guide tube in the fuel assembly.

  17. Cooling assembly for fuel cells

    DOEpatents

    Kaufman, Arthur; Werth, John

    1990-01-01

    A cooling assembly for fuel cells having a simplified construction whereby coolant is efficiently circulated through a conduit arranged in serpentine fashion in a channel within a member of such assembly. The channel is adapted to cradle a flexible, chemically inert, conformable conduit capable of manipulation into a variety of cooling patterns without crimping or otherwise restricting of coolant flow. The conduit, when assembled with the member, conforms into intimate contact with the member for good thermal conductivity. The conduit is non-corrodible and can be constructed as a single, manifold-free, continuous coolant passage means having only one inlet and one outlet.

  18. Reusable fuel test assembly for the FFTF

    SciTech Connect

    Pitner, A.L.; Dittmer, J.O. )

    1992-01-01

    A fuel test assembly that provides re-irradiation capability after interim discharge and reconstitution of the test pin bundle has been developed for use in the Fast Flux Test Facility (FFTF). This test vehicle permits irradiation test data to be obtained at multiple exposures on a few select test pins without the substantial expense of fabricating individual test assemblies as would otherwise be required. A variety of test pin types can be loaded in the reusable test assembly. A reusable test vehicle for irradiation testing in the FFTF has long been desired, but a number of obstacles previously prevented the implementation of such an experimental rig. The MFF-8A test assembly employs a 169-pin bundle using HT-9 alloy for duct and cladding material. The standard driver pins in the fuel bundle are sodium-bonded metal fuel (U-10 wt% Zr). Thirty-seven positions in the bundle are replaceable pin positions. Standard MFF-8A driver pins can be loaded in any test pin location to fill the bundle if necessary. Application of the MFF-8A reusable test assembly in the FFTF constitutes a considerable cost-saving measure with regard to irradiation testing. Only a few well-characterized test pins need be fabricated to conduct a test program rather than constructing entire test assemblies.

  19. 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.

  20. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    DOEpatents

    Stengel, F.G.

    1963-12-24

    A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

  1. Improved nuclear fuel assembly grid spacer

    DOEpatents

    Marshall, John; Kaplan, Samuel

    1977-01-01

    An improved fuel assembly grid spacer and method of retaining the basic fuel rod support elements in position within the fuel assembly containment channel. The improvement involves attachment of the grids to the hexagonal channel and of forming the basic fuel rod support element into a grid structure, which provides a design which is insensitive to potential channel distortion (ballooning) at high fluence levels. In addition the improved method eliminates problems associated with component fabrication and assembly.

  2. Precharacterization Report for Instrumented Fuel Assembly (IFA)-527

    SciTech Connect

    Cunningham, M. E.; Bradley, E. R.; Daniel, J. L.; Davis, N. C.; Lanning, D. D.; Williford, R. E.

    1981-07-01

    This report is a resource document covering the rationale, design, fabrication, and preirradiation characterization of instrumented fuel assembly (IFA)-527. This assembly is being irradiated in the Halden Boiling Water Reactor (HBWR) in Norway as part of the Experimental Support and Development of Single-Rod Fuel Codes Program conducted by Pacific Northwest laboratory (PNL) and sponsored by the Fuel Behavior Research Branch of the U.S. Nuclear Regulatory Commission (NRC). Data from this assembly will be used to better understand light water reactor (LWR) fuel behavior under normal operating conditions.

  3. Apparatus for shearing spent nuclear fuel assemblies

    DOEpatents

    Weil, Bradley S.; Metz, III, Curtis F.

    1980-01-01

    A method and apparatus are described for shearing spent nuclear fuel assemblies of the type comprising an array of fuel pins disposed within an outer metal shell or shroud. A spent fuel assembly is first compacted in a known manner and then incrementally sheared using fixed and movable shear blades having matched laterally projecting teeth which slidably intermesh to provide the desired shearing action. Incremental advancement of the fuel assembly after each shear cycle is limited to a distance corresponding to the lateral projection of the teeth to ensure fuel assembly breakup into small uniform segments which are amenable to remote chemical processing.

  4. Fuel rod assembly to manifold attachment

    DOEpatents

    Donck, Harry A.; Veca, Anthony R.; Snyder, Jr., Harold J.

    1980-01-01

    A fuel element is formed with a plurality of fuel rod assemblies detachably connected to an overhead support with each of the fuel rod assemblies having a gas tight seal with the support to allow internal fission gaseous products to flow without leakage from the fuel rod assemblies into a vent manifold passageway system on the support. The upper ends of the fuel rod assemblies are located at vertically extending openings in the support and upper threaded members are threaded to the fuel rod assemblies to connect the latter to the support. The preferred threaded members are cap nuts having a dome wall encircling an upper threaded end on the fuel rod assembly and having an upper sealing surface for sealing contact with the support. Another and lower seal is achieved by abutting a sealing surface on each fuel rod assembly with the support. A deformable portion on the cap nut locks the latter against inadvertent turning off the fuel rod assembly. Orienting means on the fuel rod and support primarily locates the fuel rods azimuthally for reception of a deforming tool for the cap nut. A cross port in the fuel rod end plug discharges into a sealed annulus within the support, which serves as a circumferential chamber, connecting the manifold gas passageways in the support.

  5. Fuel Cell Electrodes for Hydrogen-Air Fuel Cell Assemblies.

    DTIC Science & Technology

    The report describes the design and evaluation of a hydrogen-air fuel cell module for use in a portable hydrid fuel cell -battery system. The fuel ... cell module consists of a stack of 20 single assemblies. Each assembly contains 2 electrically independent cells with a common electrolyte compartment

  6. Method for shearing spent nuclear fuel assemblies

    DOEpatents

    Weil, Bradley S.; Watson, Clyde D.

    1977-01-01

    A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

  7. 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.

  8. Thermal Analysis of a TREAT Fuel Assembly

    SciTech Connect

    Papadias, Dionissios; Wright, Arthur E.

    2014-07-09

    The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

  9. Calibration of spent fuel measurement assembly

    NASA Astrophysics Data System (ADS)

    Koleska, Michal; Viererbl, Ladislav; Marek, Milan

    2014-11-01

    The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110mAg isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110mAg isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system.

  10. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    DOEpatents

    Schluderberg, D.C.; Ryon, J.W.

    1962-05-01

    A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

  11. Polymer electrolyte membrane assembly for fuel cells

    NASA Technical Reports Server (NTRS)

    Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

    2000-01-01

    An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

  12. Polymer electrolyte membrane assembly for fuel cells

    NASA Technical Reports Server (NTRS)

    Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

    2002-01-01

    An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

  13. Integrated regenerative fuel cell experimental evaluation

    NASA Technical Reports Server (NTRS)

    Martin, Ronald E.

    1990-01-01

    An experimental test program was conducted to investigate the performance characteristics of an integrated regenerative fuel cell (IRFC) concept. The IRFC consists of a separate fuel cell unit and electrolysis cell unit in the same structure, with internal storage of fuel cell product water and external storage of electrolysis cell produced hydrogen and oxygen. The fuel cell unit incorporates an enhanced Orbiter-type cell capable of improved performance at reduced weight. The electrolysis cell features a NiCo2O4 catalyst oxygen evolution eletrode with a porous Teflon cover to retard electrolyte loss. Six complete IRFC assemblies were assembled and performance tested at an operating temperature of 200 F (93.3 C) and reactant pressures up to 170 psia (117.2 n/cu cm) on IRFC No. 4. Anomalous pressure charge/discharge characteristics were encountered during performance evaluation. A reversible fuel cell incorporating a proprietary bi-functional oxygen electrode operated satisfactory at 200 F (93.3 C) at reactant pressures up to 50 psia (41.4 n/cu cm) as a regenerative fuel cell for one cycle, before developing an electrical short in the fuel cell mode. Electrolysis cell 300-hour endurance tests demonstrated the electrolyte retention capability of the electrode Teflon cover and the performance stability of the bi-functional oxygen electrode at high potential.

  14. Fuel cell with electrolyte matrix assembly

    DOEpatents

    Kaufman, Arthur; Pudick, Sheldon; Wang, Chiu L.

    1988-01-01

    This invention is directed to a fuel cell employing a substantially immobilized electrolyte imbedded therein and having a laminated matrix assembly disposed between the electrodes of the cell for holding and distributing the electrolyte. The matrix assembly comprises a non-conducting fibrous material such as silicon carbide whiskers having a relatively large void-fraction and a layer of material having a relatively small void-fraction.

  15. Fuel cell assembly with electrolyte transport

    DOEpatents

    Chi, Chang V.

    1983-01-01

    A fuel cell assembly wherein electrolyte for filling the fuel cell matrix is carried via a transport system comprising a first passage means for conveying electrolyte through a first plate and communicating with a groove in a second plate at a first point, the first and second plates together sandwiching the matrix, and second passage means acting to carry electrolyte exclusively through the second plate and communicating with the groove at a second point exclusive of the first point.

  16. Membrane electrode assembly for a fuel cell

    NASA Technical Reports Server (NTRS)

    Prakash, Surya (Inventor); Narayanan, Sekharipuram R. (Inventor); Atti, Anthony (Inventor); Olah, George (Inventor); Smart, Marshall C. (Inventor)

    2006-01-01

    A catalyst ink for a fuel cell including a catalytic material and poly(vinylidene fluoride). The ink may be applied to a substrate to form an electrode, or bonded with other electrode layers to form a membrane electrode assembly (MEA).

  17. Processing of driver fuel assemblies at FFTF

    SciTech Connect

    Danko, A.D.; Hicks, D.F.; Arneson, S.O.

    1982-07-01

    The ability to disassemble an irradiated Fast Flux Test Facility (FFTF) Driver Fuel Assembly (DFA) is important both to the continued operation of the FFTF and the future of the Breeder Reactor Program. At the FFTF, DFA's with up to three (3)* kilowatts of decay heat will be placed in the Interim Examination and Maintenance (IEM) Cell for disassembly and nondestructive examination. This process includes sodium removal, duct measurement, duct cutting and pulling, fuel pin removal, and component disposition to other laboratories for destructive examination.

  18. The Conceptual Design for a Fuel Assembly of a New Research Reactor

    SciTech Connect

    Ryu, J-S.; Cho, Y-G.; Yoon, D-B.; Dan, H-J.; Chae, H-T.; Park, C.

    2004-10-06

    A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibration characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.

  19. Inspection procedures for experimental fuel production

    NASA Astrophysics Data System (ADS)

    Campsie, I. C.; Rattray, H. D.

    1988-04-01

    This paper describes the inspection procedures used in the development and manufacture of experimental fuel elements and their components. The examples quoted mainly apply to the PFR experimental fuel programme, although for well over a quarter of a century the procedures and techniques have been progressively developed and applied to the Magnox, SGHW, AGR, HTR, PFR and PWR fuel development programmes undertaken at the UKAEA's Springfields and Windscale Nuclear Power Development Laboratories. In contrast to production runs involving large numbers of standard components, experimental fuel is often assembled from components which, while they may look alike, may have design and material variations. Thus in addition to normal batching and bonding operations, great emphasis has to be placed on dimensional inspection, material testing and the individual identification of all items, thus maintaining traceability throughout all operations. The quality and performance of experimental items are often evaluated comparing pre- and post-test dimensional or NDT measurements. In the case of irradiation tests, several years can elapse between the measurements, therefore it is essential to ensure the reproducibility and compatibility of pre- and post-test measuring techniques and the traceability of all measured data and standards.

  20. Selected Isotopes for Optimized Fuel Assembly Tags

    SciTech Connect

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2008-10-01

    In support of our ongoing signatures project we present information on 3 isotopes selected for possible application in optimized tags that could be applied to fuel assemblies to provide an objective measure of burnup. 1. Important factors for an optimized tag are compatibility with the reactor environment (corrosion resistance), low radioactive activation, at least 2 stable isotopes, moderate neutron absorption cross-section, which gives significant changes in isotope ratios over typical fuel assembly irradiation levels, and ease of measurement in the SIMS machine 2. From the candidate isotopes presented in the 3rd FY 08 Quarterly Report, the most promising appear to be Titanium, Hafnium, and Platinum. The other candidate isotopes (Iron, Tungsten, exhibited inadequate corrosion resistance and/or had neutron capture cross-sections either too high or too low for the burnup range of interest.

  1. Fuel assembly shaker and truck test simulation

    SciTech Connect

    Klymyshyn, Nicholas A.; Jensen, Philip J.; Sanborn, Scott E.; Hanson, Brady D.

    2014-09-30

    This study continues the modeling support of the SNL shaker table task from 2013 and includes analysis of the SNL 2014 truck test campaign. Detailed finite element models of the fuel assembly surrogate used by SNL during testing form the basis of the modeling effort. Additional analysis was performed to characterize and filter the accelerometer data collected during the SNL testing. The detailed fuel assembly finite element model was modified to improve the performance and accuracy of the original surrogate fuel assembly model in an attempt to achieve a closer agreement with the low strains measured during testing. The revised model was used to recalculate the shaker table load response from the 2013 test campaign. As it happened, the results remained comparable to the values calculated with the original fuel assembly model. From this it is concluded that the original model was suitable for the task and the improvements to the model were not able to bring the calculated strain values down to the extremely low level recorded during testing. The model needs more precision to calculate strains that are so close to zero. The truck test load case had an even lower magnitude than the shaker table case. Strain gage data from the test was compared directly to locations on the model. Truck test strains were lower than the shaker table case, but the model achieved a better relative agreement of 100-200 microstrains (or 0.0001-0.0002 mm/mm). The truck test data included a number of accelerometers at various locations on the truck bed, surrogate basket, and surrogate fuel assembly. This set of accelerometers allowed an evaluation of the dynamics of the conveyance system used in testing. It was discovered that the dynamic load transference through the conveyance has a strong frequency-range dependency. This suggests that different conveyance configurations could behave differently and transmit different magnitudes of loads to the fuel even when traveling down the same road at

  2. Advanced membrane electrode assemblies for fuel cells

    SciTech Connect

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  3. Advanced membrane electrode assemblies for fuel cells

    SciTech Connect

    Kim, Yu Seung; Pivovar, Bryan S.

    2012-07-24

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  4. Experimental investigations of heat transfer and temperature fields in models simulating fuel assemblies used in the core of a nuclear reactor with a liquid heavy-metal coolant

    NASA Astrophysics Data System (ADS)

    Belyaev, I. A.; Genin, L. G.; Krylov, S. G.; Novikov, A. O.; Razuvanov, N. G.; Sviridov, V. G.

    2015-09-01

    The aim of this experimental investigation is to obtain information on the temperature fields and heat transfer coefficients during flow of liquid-metal coolant in models simulating an elementary cell in the core of a liquid heavy metal cooled fast-neutron reactor. Two design versions for spacing fuel rods in the reactor core were considered. In the first version, the fuel rods were spaced apart from one another using helical wire wound on the fuel rod external surface, and in the second version spacer grids were used for the same purpose. The experiments were carried out on the mercury loop available at the Moscow Power Engineering Institute National Research University's Chair of Engineering Thermal Physics. Two experimental sections simulating an elementary cell for each of the fuel rod spacing versions were fabricated. The temperature fields were investigated using a dedicated hinged probe that allows temperature to be measured at any point of the studied channel cross section. The heat-transfer coefficients were determined using the wall temperature values obtained at the moment when the probe thermocouple tail end touched the channel wall. Such method of determining the wall temperature makes it possible to alleviate errors that are unavoidable in case of measuring the wall temperature using thermocouples placed in slots milled in the wall. In carrying out the experiments, an automated system of scientific research was applied, which allows a large body of data to be obtained within a short period of time. The experimental investigations in the first test section were carried out at Re = 8700, and in the second one, at five values of Reynolds number. Information about temperature fields was obtained by statistically processing the array of sampled probe thermocouple indications at 300 points in the experimental channel cross section. Reach material has been obtained for verifying the codes used for calculating velocity and temperature fields in channels with

  5. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    NASA Astrophysics Data System (ADS)

    Panferov, Pavel; Kochkin, Viacheslav; Erak, Dmitry; Makhotin, Denis; Reshetnikov, Alexandr; Timofeev, Andrey

    2016-02-01

    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  6. Pressurized water reactor fuel assembly subchannel void fraction measurement

    SciTech Connect

    Akiyama, Yoshiei; Hori, Keiichi; Miyazaki, Keiji; Mishima, Kaichiro; Sugiyama, Shigekazu

    1995-12-01

    The void fraction measurement experiment of pressurized water reactor (PWR) fuel assemblies has been conducted since 1987 under the sponsorship of the Ministry of International Trade and Industry as a Japanese national project. Two types of test sections are used in this experiment. One is a 5 x 5 array rod bundle geometry, and the other is a single-channel geometry simulating one of the subchannels in the rod bundle. Wide gamma-ray beam scanners and narrow gamma-ray beam computed tomography scanners are used to measure the subchannel void fractions under various steady-state and transient conditions. The experimental data are expected to be used to develop a void fraction prediction model relevant to PWR fuel assemblies and also to verify or improve the subchannel analysis method. The first series of experiments was conducted in 1992, and a preliminary evaluation of the data has been performed. The preliminary results of these experiments are described.

  7. Interface ring for gas turbine fuel nozzle assemblies

    SciTech Connect

    Fox, Timothy A.; Schilp, Reinhard

    2016-03-22

    A gas turbine combustor assembly including a combustor liner and a plurality of fuel nozzle assemblies arranged in an annular array extending within the combustor liner. The fuel nozzle assemblies each include fuel nozzle body integral with a swirler assembly, and the swirler assemblies each include a bellmouth structure to turn air radially inwardly for passage into the swirler assemblies. A radially outer removed portion of each of the bellmouth structures defines a periphery diameter spaced from an inner surface of the combustor liner, and an interface ring is provided extending between the combustor liner and the removed portions of the bellmouth structures at the periphery diameter.

  8. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    SciTech Connect

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-12-31

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR`s) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design.

  9. Fuel injection assembly for use in turbine engines and method of assembling same

    SciTech Connect

    Berry, Jonathan Dwight; Johnson, Thomas Edward; York, William David; Uhm, Jong Ho

    2015-12-15

    A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes an end cover, an endcap assembly, a fluid supply chamber, and a plurality of tube assemblies positioned at the endcap assembly. Each of the tube assemblies includes housing having a fuel plenum and a cooling fluid plenum. The cooling fluid plenum is positioned downstream from the fuel plenum and separated from the fuel plenum by an intermediate wall. The plurality of tube assemblies also include a plurality of tubes that extends through the housing. Each of the plurality of tubes is coupled in flow communication with the fluid supply chamber and a combustion chamber positioned downstream from the tube assembly. The plurality of tube assemblies further includes an aft plate at a downstream end of the cooling fluid plenum. The plate includes at least one aperture.

  10. Cross flow characteristics in a three fuel assemblies

    SciTech Connect

    Bae, J. H.; Euh, D. J.; Park, C. K.; Youn, Y. J.; Kwon, T. S.

    2012-07-01

    To evaluate the reactor thermal margin of APR+, reactor core flow distribution including both axial and lateral directional hydraulic resistances of fuel assemblies should be known. 3-Ch cross flow test facility has been constructed with three full-size fuel assemblies to investigate the cross flow characteristics. Performance tests have been performed. The axial and lateral directional hydraulic resistances of fuel assemblies have been measured. The test results have been compared to the CFD calculation. (authors)

  11. Fuel fire tests of selected assemblies. Interim report

    SciTech Connect

    Kydd, G.; Spindola, K.; Askew, G.K.

    1982-04-13

    A varing assortment of clothing assemblies was tested in the Fuel Fire Test Facility at the Naval Air Development Center. Included was a Nomex-Kevlar Cloque Coverall which had relatively good protection from fuel flames.

  12. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    DOE PAGES

    Mertyurek, Ugur; Gauld, Ian C.

    2015-12-24

    In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

  13. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    SciTech Connect

    Mertyurek, Ugur; Gauld, Ian C.

    2015-12-24

    In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

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

    SciTech Connect

    Canaan, R.E.

    1995-12-01

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

  15. Reversible BWR fuel assembly and method of using same

    SciTech Connect

    Freeman, T.R.; Wilson, J.F.; Knott, R.P.

    1987-04-07

    A nuclear fuel assembly is described comprising: (a) a flow channel; (b) a lower nozzle assembly structurally attached to the flow channel to form therewith an external envelope; (c) an invertible fuel bundle adapted to be inserted into the envelope, the fuel bundle comprising elongated fuel rods held in a spaced lateral array between top and bottom tie plates. Each of the top and bottom tie plates is substantially identical and has means for supporting the fuel bundle within the envelope in either of two mutually inverted vertical orientations whereby the orientation of the fuel bundle in a flow channel may be reversed during burn-up operation.

  16. Separator assembly for use in spent nuclear fuel shipping cask

    DOEpatents

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  17. Mechanical Analysis of the Fuel Assembly Box of a HPLWR Fuel Assembly

    SciTech Connect

    Himmel, Steffen; Starflinger, Joerg; Schulenberg, Thomas; Hofmeister, Jan

    2006-07-01

    The aim of the work presented in this paper is to demonstrate that the assembly box of the fuel assembly for a HPLWR proposed by Hofmeister et al. will remain mechanically within the design limits. The commercial finite element code ANSYS has been used to investigate the deformation behaviour caused by thermal convective and pressure boundary conditions provided by the results from Waata et al. for the fuel assembly. The results of these ANSYS analyses show a bending of the assembly box caused by the applied temperature and pressure distribution which, however, is still within the geometrical allowances. The maximum bending of the 4.35 m long assembly box appears close to the mid section, i.e. at 2.45 m axial height, and amounts to about 2 mm, only. The maximum indentation is mainly caused by the pressure difference across the box wall and occurs near the top of the assembly. The indentation at this point can be evaluated to be around 0.2 mm. Both bending and indentation will influence the coolant mass flux and the moderator distribution, and thus needs to be considered for predictions of the power profile and of the coolant heat-up. They are not considered to be critical as long as these deformations are small compared with the nominal gap width of 1 mm between box wall and claddings and 10 mm between adjacent assembly boxes. A second analysis has been performed to study the effect on non-symmetric coolant temperature profiles. A coolant temperature increase by 30 deg. C on one side of the box increased the thermal bending to 4 mm, indicating the sensitivity of this design with respect to temperature non-uniformities. (authors)

  18. Some methods for achieving more efficient performance of fuel assemblies

    NASA Astrophysics Data System (ADS)

    Boltenko, E. A.

    2014-07-01

    More efficient operation of reactor plant fuel assemblies can be achieved through the use of new technical solutions aimed at obtaining more uniform distribution of coolant over the fuel assembly section, more intense heat removal on convex heat-transfer surfaces, and higher values of departure from nucleate boiling ratio (DNBR). Technical solutions using which it is possible to obtain more intense heat removal on convex heat-transfer surfaces and higher DNBR values in reactor plant fuel assemblies are considered. An alternative heat removal arrangement is described using which it is possible to obtain a significantly higher power density in a reactor plant and essentially lower maximal fuel rod temperature.

  19. Fuel assembly cooling experience at the FFTF IEM cell

    SciTech Connect

    McGuinness, P.W.

    1985-11-01

    To date, 13 fuel assemblies requiring forced cooling have been processed through the Fast Flux Test Facility (FFTF) interim examination and maintenance (IEM) cell. Of these, two assemblies experienced overtemperature conditions due to inadequate forced cooling. Both of the occurrences have contributed significantly to the process of learning how to operate a fuel assembly cooling system remotely in an argon atmosphere hot cell. Many innovations have been made to the cooling system to enhance safety and increase productivity, and are briefly described.

  20. Mechanical Tests and Analyses on the FBR Ductless Fuel Assembly

    SciTech Connect

    Itoh, K.; Sato, T.; Ogura, M.; Ohkubo, Y.; Moro, S.; Madarame, H.

    2002-07-01

    Fast Breeder Reactor (FBR) cores, which are composed of ductless fuel assemblies, have many potentialities of cost reduction through whole fuel cycle procedures with safety features, and of reduced high-level-waste disposal. The mechanistic aspects of the ductless fuel assembly have been investigated by mechanical tests and analyses, and the core static behaviors under the irradiation and the seismic occasion have been clarified in this study. (authors)

  1. Method and apparatus for close packing of nuclear fuel assemblies

    DOEpatents

    Newman, Darrell F.

    1993-01-01

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  2. Method and apparatus for close packing of nuclear fuel assemblies

    DOEpatents

    Newman, D.F.

    1993-03-30

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  3. Criticality safety evaluation report for FFTF 42% fuel assemblies

    SciTech Connect

    Richard, R.F.

    1997-10-28

    An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC).

  4. Combustor with two stage primary fuel assembly

    DOEpatents

    Sharifi, Mehran; Zolyomi, Wendel; Whidden, Graydon Lane

    2000-01-01

    A combustor for a gas turbine having first and second passages for pre-mixing primary fuel and air supplied to a primary combustion zone. The flow of fuel to the first and second pre-mixing passages is separately regulated using a single annular fuel distribution ring having first and second row of fuel discharge ports. The interior portion of the fuel distribution ring is divided by a baffle into first and second fuel distribution manifolds and is located upstream of the inlets to the two pre-mixing passages. The annular fuel distribution ring is supplied with fuel by an annular fuel supply manifold, the interior portion of which is divided by a baffle into first and second fuel supply manifolds. A first flow of fuel is regulated by a first control valve and directed to the first fuel supply manifold, from which the fuel is distributed to first fuel supply tubes that direct it to the first fuel distribution manifold. From the first fuel distribution manifold, the first flow of fuel is distributed to the first row of fuel discharge ports, which direct it into the first pre-mixing passage. A second flow of fuel is regulated by a second control valve and directed to the second fuel supply manifold, from which the fuel is distributed to second fuel supply tubes that direct it to the second fuel distribution manifold. From the second fuel distribution manifold, the second flow of fuel is distributed to the second row of fuel discharge ports, which direct it into the second pre-mixing passage.

  5. Fuel injection assembly for use in turbine engines and method of assembling same

    DOEpatents

    Uhm, Jong Ho; Johnson, Thomas Edward

    2015-03-24

    A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes a plurality of tube assemblies, wherein each of the tube assemblies includes an upstream portion and a downstream portion. Each tube assembly includes a plurality of tubes that extend from the upstream portion to the downstream portion or from the upstream portion through the downstream portion. At least one injection system is coupled to at least one tube assembly of the plurality of tube assemblies. The injection system includes a fluid supply member that extends from a fluid source to the downstream portion of the tube assembly. The fluid supply member includes a first end portion located in the downstream portion of the tube assembly, wherein the first end portion has at least one first opening for channeling fluid through the tube assembly to facilitate reducing a temperature therein.

  6. Advanced fuel assembly characterization capabilities based on gamma tomography at the Halden boiling water reactor

    SciTech Connect

    Holcombe, S.; Eitrheim, K.; Svaerd, S. J.; Hallstadius, L.; Willman, C.

    2012-07-01

    Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Inst. for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala Univ.. (authors)

  7. Fuel burner and combustor assembly for a gas turbine engine

    DOEpatents

    Leto, Anthony

    1983-01-01

    A fuel burner and combustor assembly for a gas turbine engine has a housing within the casing of the gas turbine engine which housing defines a combustion chamber and at least one fuel burner secured to one end of the housing and extending into the combustion chamber. The other end of the fuel burner is arranged to slidably engage a fuel inlet connector extending radially inwardly from the engine casing so that fuel is supplied, from a source thereof, to the fuel burner. The fuel inlet connector and fuel burner coact to anchor the housing against axial movement relative to the engine casing while allowing relative radial movement between the engine casing and the fuel burner and, at the same time, providing fuel flow to the fuel burner. For dual fuel capability, a fuel injector is provided in said fuel burner with a flexible fuel supply pipe so that the fuel injector and fuel burner form a unitary structure which moves with the fuel burner.

  8. Temperature measuring analysis of the nuclear reactor fuel assembly

    SciTech Connect

    Urban, F. E-mail: zdenko.zavodny@stuba.sk; Kučák, L. E-mail: zdenko.zavodny@stuba.sk; Bereznai, J. E-mail: zdenko.zavodny@stuba.sk; Závodný, Z. E-mail: zdenko.zavodny@stuba.sk; Muškát, P. E-mail: zdenko.zavodny@stuba.sk

    2014-08-06

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  9. Current conducting end plate of fuel cell assembly

    DOEpatents

    Walsh, Michael M.

    1999-01-01

    A fuel cell assembly has a current conducting end plate with a conductive body formed integrally with isolating material. The conductive body has a first surface, a second surface opposite the first surface, and an electrical connector. The first surface has an exposed portion for conducting current between a working section of the fuel cell assembly and the electrical connector. The isolating material is positioned on at least a portion of the second surface. The conductive body can have support passage(s) extending therethrough for receiving structural member(s) of the fuel cell assembly. Isolating material can electrically isolate the conductive body from the structural member(s). The conductive body can have service passage(s) extending therethrough for servicing one or more fluids for the fuel cell assembly. Isolating material can chemically isolate the one or more fluids from the conductive body. The isolating material can also electrically isolate the conductive body from the one or more fluids.

  10. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly.

    PubMed

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-04-21

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  11. Fuel assembly cooling experience at the FFTF/IEM cell

    SciTech Connect

    McGuinness, P.W.

    1985-01-01

    In the Fast Flux Test Facility (FFTF), sodium wetted irradiated fuel assemblies are discharged to the Interim Examination and Maintenance (IEM) Cell for disassembly and post-irradiation examination in an inert argon atmosphere. While in the IEM Cell, fuel assemblies are cooled by the IEM Cell Subassembly Cooling System. This paper describes the cooling system design, performance, and lessons learned, including a discussion of two overtemperature incidents. 2 refs., 6 figs.

  12. Portable instrument for inspecting irradiated nuclear fuel assemblies

    DOEpatents

    Nicholson, Nicholas; Dowdy, Edward J.; Holt, David M.; Stump, Jr., Charles J.

    1985-01-01

    A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

  13. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    DOEpatents

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  14. PWR and BWR spent fuel assembly gamma spectra measurements

    NASA Astrophysics Data System (ADS)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  15. PWR and BWR spent fuel assembly gamma spectra measurements

    SciTech Connect

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Hu, Jianwei; Schwalbach, P.; Sjoland, A.; Trellue, Holly; Vo, D.

    2016-07-17

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  16. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGES

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

    2016-07-17

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  17. Hydrogen storage and integrated fuel cell assembly

    DOEpatents

    Gross, Karl J.

    2010-08-24

    Hydrogen is stored in materials that absorb and desorb hydrogen with temperature dependent rates. A housing is provided that allows for the storage of one or more types of hydrogen-storage materials in close thermal proximity to a fuel cell stack. This arrangement, which includes alternating fuel cell stack and hydrogen-storage units, allows for close thermal matching of the hydrogen storage material and the fuel cell stack. Also, the present invention allows for tailoring of the hydrogen delivery by mixing different materials in one unit. Thermal insulation alternatively allows for a highly efficient unit. Individual power modules including one fuel cell stack surrounded by a pair of hydrogen-storage units allows for distribution of power throughout a vehicle or other electric power consuming devices.

  18. Storage assembly for spent nuclear fuel

    SciTech Connect

    Lapides, M.E.

    1982-04-27

    A technique for storing spent fuel rods from a nuclear reactor is disclosed herein. This technique utilizes a housing including a closed inner chamber for containing the fuel rods and a thermally conductive member located partially within the housing chamber and partially outside the housing for transferring heat generated by the fuel rods from the chamber to the ambient surroundings. Particulate material is located within the chamber and surrounds the fuel rods contained therein. This material is selected to serve as a heat transfer media between the contained cells and the heat transferring member and, at the same time, stand ready to fuse into a solid mass around the contained cells if the heat transferring member malfunctions or otherwise fails to transfer the generated heat out of the housing chamber in a predetermined way.

  19. Analysis of subcritical experiments using fresh and spent research reactor fuel assemblies

    NASA Astrophysics Data System (ADS)

    Zino, John Frederick

    1999-11-01

    This research investigated the concepts associated with crediting the burnup of spent nuclear fuel assemblies for the purposes of criticality safety. To accomplish this, a collaborative experimental research program was undertaken between Westinghouse, the University of Missouri Research Reactor (MURR) facility and Oak Ridge National Laboratory (ORNL). The purpose of the program was to characterize the subcritical behavior of a small array of fresh and spent MURR fuel assemblies using the 252Cf Source-driven noise technique. An aluminum test rig was built which was capable of holding up to four, highly enriched (93.15 wt.% 235U) MURR fuel assemblies in a 2 x 2 array. The rig was outfitted with one source and four detector drywells which allowed researchers to perform active neutron noise measurements on the array of fuel assemblies. The 1 atmosphere gas 3He neutron detectors used to perform the measurements were quenched with CF4 gas to allow improved discrimination of the neutron signals in the very high gamma-ray fields associated with spent fuel (˜8000 R/hr). In addition, the detector drywells were outfitted with 1″ lead collars to provide additional gamma-ray shielding from the spent fuel. Reactivity changes were induced in the subcritical lattice by replacing individual fresh assemblies (in a 4-assembly array) with spent assemblies of known, maximum burnup (143 Mw-D). The absolute and relative measured reactivity changes were then compared to those predicted by three-dimensional Monte Carlo calculations. The purpose of these comparisons was to investigate the accuracy of modern transport theory depletion calculations to accurately simulate the reactivity effects of burnup in spent nuclear fuel. A total of seven subcritical measurements were performed at the MURR reactor facility on July 20th and 27th, 1998. These measurements generated several estimates of prompt neutron decay constants (alpha) and ratios of spectral densities through frequency correlations

  20. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    SciTech Connect

    Kyser, E.

    2010-06-17

    A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

  1. Storage, transportation and disposal system for used nuclear fuel assemblies

    DOEpatents

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  2. Fuel assembly duct cutting in the FFTF/IEM Cell

    SciTech Connect

    Gibbons, P.W.

    1985-01-01

    Two mill type slitting cutters are used in the Fast Flux Test Facility (FFTF) Interim Examination and Maintenance (IEM) Cell during the disassembly sequence of a Driver Fuel Assembly. This disassembly is necessary so that selected parts may be examined both in the IEM Cell and elsewhere. The cutters have been in use for two years. During this time eight Driver Fuel assemblies have been taken apart in the IEM Cell. The cutters' operating philosophy and characteristics, as well as lessons learned from a significant equipment failure are presented. 1 ref., 6 figs., 1 tab.

  3. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1986-01-01

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engageable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  4. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.

    1983-08-29

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  5. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    NASA Astrophysics Data System (ADS)

    Jacobsson Svärd, Staffan; Holcombe, Scott; Grape, Sophie

    2015-05-01

    A fuel assembly operated in a nuclear power plant typically contains 100-300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative assessment, which

  6. EXPERIMENTAL LIQUID METAL FUEL REACTOR

    DOEpatents

    Happell, J.J.; Thomas, G.R.; Denise, R.P.; Bunts, J.L. Jr.

    1962-01-23

    A liquid metal fuel nuclear fission reactor is designed in which the fissionable material is dissolved or suspended in a liquid metal moderator and coolant. The liquid suspension flows into a chamber in which a critical amount of fissionable material is obtained. The fluid leaves the chamber and the heat of fission is extracted for power or other utilization. The improvement is in the support arrangement for a segrnented graphite core to permit dif ferential thermal expansion, effective sealing between main and blanket liquid metal flows, and avoidance of excessive stress development in the graphite segments. (AEC)

  7. Method and apparatus for assembling solid oxide fuel cells

    DOEpatents

    Szreders, B.E.; Campanella, N.

    1988-05-11

    This invention relates generally to solid oxide fuel power generators and is particularly directed to improvements in the assembly and coupling of solid oxide fuel cell modules. A plurality of jet air tubes are supported and maintained in a spaced matrix array by a positioning/insertion assembly for insertion in respective tubes of a solid oxide fuel cell (SOFC) in the assembly of an SOFC module. The positioning/insertion assembly includes a plurality of generally planar, elongated, linear vanes which are pivotally mounted at each end thereof to a support frame. A rectangular compression assembly of adjustable size is adapted to receive and squeeze a matrix of SOFC tubes so as to compress the inter-tube nickel felt conductive pads which provide series/parallel electrical connection between adjacent SOFCs, with a series of increasingly larger retainer frames used to maintain larger matrices of SOFC tubes in position. Expansion of the SOFC module housing at the high operating temperatures of the SOFC is accommodated by conductive, flexible, resilient expansion, connector bars which provide support and electrical coupling at the top and bottom of the SOFC module housing. 17 figs.

  8. Fuel-rich, catalytic reaction experimental results

    NASA Technical Reports Server (NTRS)

    Rollbuhler, R. James

    1991-01-01

    Future aeropropulsion gas turbine combustion requirements call for operating at very high inlet temperatures, pressures, and large temperature rises. At the same time, the combustion process is to have minimum pollution effects on the environment. Aircraft gas turbine engines utilize liquid hydrocarbon fuels which are difficult to uniformly atomize and mix with combustion air. An approach for minimizing fuel related problems is to transform the liquid fuel into gaseous form prior to the completion of the combustion process. Experimentally obtained results are presented for vaporizing and partially oxidizing a liquid hydrocarbon fuel into burnable gaseous components. The presented experimental data show that 1200 to 1300 K reaction product gas, rich in hydrogen, carbon monoxide, and light-end hydrocarbons, is formed when flowing 0.3 to 0.6 fuel to air mixes through a catalyst reactor. The reaction temperatures are kept low enough that nitrogen oxides and carbon particles (soot) do not form. Results are reported for tests using different catalyst types and configurations, mass flowrates, input temperatures, and fuel to air ratios.

  9. Experimental study of external fuel vaporization

    NASA Technical Reports Server (NTRS)

    Szetela, E. J.; Tevelde, J. A.

    1982-01-01

    The fuel properties used in the design of a flash vaporization system for aircraft gas turbine engines were evaluated in experiments using a flowing system to determine critical temperature and pressure, boiling points, dew points, heat transfer coefficients, deposit formation rates, and deposit removal. Three fuels were included in the experiments: Jet-A, an experimental referree broad specification fuel, and a premium No. 2 diesel fuel. Engine conditions representing a NASA Energy Efficient Engine at sea-level take-off, cruise, and idle were simulated in the vaporization system and it was found that single phase flow was maintained in the heat exchanger and downstream of the throttle. Deposits encountered in the heat exchanger represented a thermal resistance as high as 1300 sq M K/watt and a deposit formation rate over 1000 gC/sq cm hr.

  10. Physical characteristics of GE (General Electric) BWR (boiling-water reactor) fuel assemblies

    SciTech Connect

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs.

  11. Photon dose rates from spent fuel assemblies with relation to self-protection (Rev. 1)

    SciTech Connect

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

    1996-02-01

    Photon dose rates as a function of fission product decay times have been calculated for spent fuel assemblies typical of MTR-type research and test reactors. Based upon these dose rates, the length of time that a spent fuel assembly will be self-protecting (dose rate greater than 100 rem/h at 1 m in air) can be estimated knowing the mass of fuel burned, the fraction of fuel burned, and the fuel assembly specific power density.

  12. Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

    SciTech Connect

    Troy Unruh; Michael Reichenberger; Phillip Ugorowski

    2013-09-01

    Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

  13. Modelling an experimental methane fuel processor

    NASA Astrophysics Data System (ADS)

    Lin, Shi-Tin; Chen, Yih-Hang; Yu, Cheng-Ching; Liu, Yen-Chun; Lee, Chiou-Hwang

    Steady-state models are developed to describe an experimental methane fuel processor that is intended to provide hydrogen for a fuel cell system for power generation (2-3 kW). First-principle reactor models are constructed to describe a series of reactions, i.e., steam and autothermal reforming (SR/ATR), high- and low-temperature water-gas shift (HTS/LTS) reactions and preferential oxidation (PROX) reactions, at different sectors of the reactor system for methane reforming as well as gas cleaning. The pre-exponential factors of the rate constants are adjusted to fit the experimental data and the resultant reactor model provides a reasonably good description of steady-state behaviour. Next, sensitivity analyses are performed to locate the optimum operating point of the fuel processor. The objective function of the optimization is fuel processor efficiency. The dominating optimization variables include: the ratios of water and oxygen to the hydrocarbon feed to the autothermal reforming reactor and the inlet temperature of the reactor. The results indicate that further improvement in fuel processor efficiency can be made with a reliable process model.

  14. Fail-safe storage rack for irradiated fuel rod assemblies

    DOEpatents

    Lewis, Donald R.

    1993-01-01

    A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

  15. Fail-safe storage rack for irradiated fuel rod assemblies

    DOEpatents

    Lewis, D.R.

    1993-03-23

    A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

  16. Premixer assembly for mixing air and fuel for combustion

    SciTech Connect

    York, William David; Johnson, Thomas Edward; Keener, Christopher Paul

    2016-12-13

    A premixer assembly for mixing air and fuel for combustion includes a plurality of tubes disposed at a head end of a combustor assembly. Also included is a tube of the plurality of tubes, the tube including an inlet end and an outlet end. Further included is at least one non-circular portion of the tube extending along a length of the tube, the at least one non-circular portion having a non-circular cross-section, and the tube having a substantially constant cross-sectional area along its length

  17. Fail-safe storage rack for fuel rod assemblies

    SciTech Connect

    Lewis, D.R.

    1991-12-31

    This report discusses a fail-safe storage rack which is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

  18. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    SciTech Connect

    Carvo, Alan E.

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  19. Fuel Assembly Calculations Using the Method of Discrete Ordinates

    SciTech Connect

    Pautz, Andreas; Langenbuch, Siegfried

    2005-02-15

    The discrete ordinates code DORT is employed to treat pin cell and fuel assembly configurations in two spatial dimensions. Despite DORT's restriction to regular (i.e., Cartesian) coordinates, we demonstrate its ability to calculate accurate pin power distributions and eigenvalues of typical reactor fuel lattices. Several numerical experiments have been performed to investigate the effects of spatial, angular, and energy discretization and to quantify their impact on the results. DORT is also used to homogenize and collapse cross-section sets within the framework of the coupled transport/burnup code system KENOREST.

  20. LWR fuel assembly designs for the transmutation of LWR Spent Fuel TRU with FCM and UO{sub 2}-ThO{sub 2} Fuels

    SciTech Connect

    Bae, G.; Hong, S. G.

    2013-07-01

    In this paper, transmutation of transuranic (TRU) nuclides from LWR spent fuels is studied by using LWR fuel assemblies which consist of UO{sub 2}-ThO{sub 2} fuel pins and FCM (Fully Ceramic Microencapsulated) fuel pins. TRU from LWR spent fuel is loaded in the kernels of the TRISO particle fuels of FCM fuel pins. In the FCM fuel pins, the TRISO particle fuels are distributed in SiC matrix having high thermal conductivity. The loading patterns of fuel pins and the fuel compositions are searched to have high transmutation rate and feasible neutronic parameters including pin power peaking, temperature reactivity coefficients, and cycle length. All studies are done only in fuel assembly calculation level. The results show that our fuel assembly designs have good transmutation performances without multi-recycling and without degradation of the safety-related neutronic parameters. (authors)

  1. COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

    SciTech Connect

    Rhodes, C.A.

    1984-12-01

    This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental

  2. COBRA-SFS predictions of single assembly spent fuel heat transfer data

    SciTech Connect

    Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.; Rector, D.R.

    1986-04-01

    The study reported here is one of several efforts to evaluate and qualify the COBRA-SFS computer code for use in spent fuel storage system thermal analysis. The ability of COBRA-SFS to predict the thermal response of two single assembly spent fuel heat transfer tests was investigated through comparisons of predictions with experimental test data. From these comparisons, conclusions regarding the computational treatment of the physical phenomena occurring within a storage system can be made. This objective was successfully accomplished as reasonable agreement between predictions and data were obtained for the 21 individual test cases of the two experiments.

  3. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    SciTech Connect

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.

    1997-10-01

    The Solution High-Energy Burst Assembly (SHEBA) was originally constructed during 1980 and was designed to be a clean free-field geometry, right-circular, cylindrically symmetric critical assembly employing U(5%)O{sub 2}F{sub 2} solution as fuel. A second version of SHEBA, employing the same fuel but equipped with a fuel pump and shielding pit, was commissioned in 1993. This report includes data and operating experience for the 1993 SHEBA only. Solution-fueled benchmark work focused on the development of experimental measurements of the characterization of SHEBA; a summary of the results are given. A description of the system and the experimental results are given in some detail in the report. Experiments were designed to: (1) study the behavior of nuclear excursions in a low-enrichment solution, (2) evaluate accidental criticality alarm detectors for fuel-processing facilities, (3) provide radiation spectra and dose measurements to benchmark radiation transport calculations on a low-enrichment solution system similar to centrifuge enrichment plants, and (4) provide radiation fields to calibrate personnel dosimetry. 15 refs., 37 figs., 10 tabs.

  4. Development of an ultrasonic cleaning method for fuel assemblies

    SciTech Connect

    Heki, H.; Komura, S.; Kato, H.; Sakai, H. ); Hattori, T. )

    1991-01-01

    Almost all radiation buildup in light water reactors is the result of the deposition of activated corrosion and wear products in out-of-core areas. After operation, a significant quantity of corrosion and wear products is deposited on the fuel rods as crud. At refueling shutdowns, these activation products are available for removal. If they can be quickly and easily removed, buildup of radioactivity on out-of-core surfaces and individual exposure dose can be greatly reduced. After studying various physical cleaning methods (e.g., water jet and ultrasonic), the ultrasonic cleaning method was selected as the most effective for fuel assembly cleaning. The ultrasonic cleaning method is especially able to efficiently clean the fuel without removing the channel box. The removed crud in the channel box would be swept out to the filtration unit. Parameter survey tests were carried out to evaluate the optimum conditions for ultrasonic cleaning using a mock-up of a short section of fuel assembly with the channel box. The ultrasonic device used was a 600-W ultrasonic transducer operating at 26-kHz ultrasonic frequency.

  5. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    PubMed Central

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  6. Control assembly for controlling a fuel cell system during shutdown and restart

    DOEpatents

    Venkataraman, Ramki; Berntsen, George; Carlson, Glenn L.; Farooque, Mohammad; Beachy, Dan; Peterhans, Stefan; Bischoff, Manfred

    2010-06-15

    A fuel cell system and method in which the fuel cell system receives and an input oxidant gas and an input fuel gas, and in which a fuel processing assembly is provided and is adapted to at least humidify the input fuel gas which is to be supplied to the anode of the fuel cell of the system whose cathode receives the oxidant input gas via an anode oxidizing assembly which is adapted to couple the output of the anode of the fuel cell to the inlet of the cathode of the fuel cell during normal operation, shutdown and restart of the fuel cell system, and in which a control assembly is further provided and is adapted to respond to shutdown of the fuel cell system during which input fuel gas and input oxidant gas cease to be received by the fuel cell system, the control assembly being further adapted to, when the fuel cell system is shut down: control the fuel cell system so as to enable a purging gas to be able to flow through the fuel processing assembly to remove humidified fuel gas from the processing assembly and to enable a purging gas to be able to flow through the anode of the fuel cell.

  7. Fuel assembly design for APR1400 with low CBC

    SciTech Connect

    Hah, Chang Joo

    2015-04-29

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to Δk{sub TARGET}. A set of new designed fuel assembly satisfies an objective function, min [f=∑{sub i}(Δk{sub FA}−Δk{sub i})], and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to Δk{sub TARGET} as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  8. Inactive end cell assembly for fuel cells for improved electrolyte management and electrical contact

    DOEpatents

    Yuh, Chao-Yi; Farooque, Mohammad; Johnsen, Richard

    2007-04-10

    An assembly for storing electrolyte in a carbonate fuel cell is provided. The combination of a soft, compliant and resilient cathode current collector and an inactive anode part including a foam anode in each assembly mitigates electrical contact loss during operation of the fuel cell stack. In addition, an electrode reservoir in the positive end assembly and an electrode sink in the negative end assembly are provided, by which ribbed and flat cathode members inhibit electrolyte migration in the fuel cell stack.

  9. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 1, Activation measurements and comparison with calculations for spent fuel assembly hardware

    SciTech Connect

    Luksic, A.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1. 5 refs., 4 figs., 21 tabs.

  10. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    SciTech Connect

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  11. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    SciTech Connect

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  12. A vented inverted fuel assembly design for an SFR

    SciTech Connect

    Vitillo, F.; Todreas, N. E.; Driscoll, M. J.

    2012-07-01

    The inverted geometry (fuel outside coolant tubes) has been previously investigated at MIT for application in gas-cooled fast reactors and pressurized water-cooled thermal reactors. Venting has also been studied for conventional fuel pins and was employed for those in the Dounreay Fast Reactor. In the present work the inverted fuel approach was adopted because it allows high fuel volume fraction, reduction of the coolant void reactivity, neutron leakage and enrichment, as well as lower pressure drop for the same channel length because grids and wire wraps are no longer necessary. Furthermore most results also apply to venting of conventional fuel pins. Physical and chemical behavior of volatile fission products in sodium was investigated to determine the maximum activity inventory which would eventually be released into the primary sodium. Results of this analysis show that the most troublesome radionuclides in terms of propensity to escape from the venting system are noble gases ({sup 85}Kr and {sup 133}Xe), and cesium ({sup 134}Cs and {sup 137}Cs). A final vented inverted fuel assembly design is proposed which meets all the design goals which have been set. Additionally purification systems were devised to reduce radionuclide activity of the coolant and the cover gas to tolerable levels. It is concluded that vented inverted (or vented conventional pin) fuel is a feasible concept and has sufficiently promising advantages - increasing fuel volume fraction to 50% and core outlet temperature by 20 deg. C, hence incrementing plant thermal efficiency by about 1% - to warrant serious consideration for future SFR designs. (authors)

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

    SciTech Connect

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

    1996-12-31

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

  14. Thermal hydraulic analysis of annular fuel-based assemblies

    SciTech Connect

    Kyu Hyun Han; Soon Heung Chang

    2004-07-01

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal subchannels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner subchannels of the seed pins, mass fluxes were high due to the grid form losses in the outer subchannels. About 43% of the heat generated from the seed pin flowed into the inner subchannel and the rest into the outer subchannel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner subchannels, temperatures and enthalpies were higher in the inner subchannels. (authors)

  15. Method and apparatus for assembling solid oxide fuel cells

    DOEpatents

    Szreders, Bernard E.; Campanella, Nicholas

    1989-01-01

    A plurality of jet air tubes are supported and maintained in a spaced matrix array by a positioning/insertion assembly for insertion in respective tubes of a solid oxide fuel cell (SOFC) in the assembly of an SOFC module. The positioning/insertion assembly includes a plurality of generally planar, elongated, linear vanes which are pivotally mounted at each end thereof to a support frame. The vanes, which each include a plurality of spaced slots along the facing edges thereof, may be pivotally displaced from a generally vertical orientation, wherein each jet air tube is positioned within and engaged by the aligned slots of a plurality of paired upper and lower vanes to facilitate their insertion in respective aligned SOFC tubes arranged in a matrix array, to an inclined orientation, wherein the jet air tubes may be removed from the positioning/insertion assembly after being inserted in the SOFC tubes. A rectangular compression assembly of adjustable size is adapted to receive and squeeze a matrix of SOFC tubes so as to compress the inter-tube nickel felt conductive pads which provide series/parallel electrical connection between adjacent SOFCs, with a series of increasingly larger retainer frames used to maintain larger matrices of SOFC tubes in position. Expansion of the SOFC module housing at the high operating temperatures of the SOFC is accommodated by conductive, flexible, resilient expansion, connector bars which provide support and electrical coupling at the top and bottom of the SOFC module housing.

  16. Experimental benchmark of MCNPX calculations against self-interrogation neutron resonance densitometry (SINRD) fresh fuel measurements

    SciTech Connect

    Menlove, Howard O; Swinhoe, Martyn T; La Fleur, Adrienne M; Charlton, William S; Lee, S Y; Tobin, S J

    2010-01-01

    We have investigated the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U concentration in a PWR 15 x 15 fresh LEU fuel assembly in air. Different measurement configurations were simulated in Monte Carlo N-Particle eXtended transport code (MCNPX) and benchmarked against experimental results. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,j) reaction peaks in fission chamber. Due to the low spontaneous fission rate of {sup 238}U (i.e. no curium in the fresh fuel), {sup 252}Cf sources were used to self-interrogate the fresh fuel pins. The resonance absorption of these neutrons in the fresh fuel pins can be measured using {sup 235}U fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the number of unknowns we are trying measure because the neutron source strength and detector-fuel assembly coupling cancel in the ratios. The agreement between MCNPX results and experimental measurements confirms the accuracy of the MCNPX models used. The development of SINRD to measure the fissile content in spent fuel is important to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in LWR spent fuel in water.

  17. Cerium migration during PEM fuel cell assembly and operation

    SciTech Connect

    Baker, Andrew M.; Torraco, Dennis; Judge, Elizabeth J.; Spernjak, Dusan; Mukundan, Rangachary; Borup, Rod L.; Advani, Suresh G.; Prasad, Ajay K.

    2015-09-14

    Cerium migration between PEM fuel cell components is influenced by potential-driven mobility, ionic diffusion, and gradients in water content. These factors were investigated in ex situ experiments and in operating fuel cells. Potential-induced migration was measured ex situ in hydrated window cells. Cerium-containing MEAs were also fabricated and tested under ASTs. MEA disassembly and subsequent XRF analysis were used to observe rapid cerium migration during cell assembly and operation. During MEA hot pressing, humidification, and low RH operation at OCV, ionic diffusion causes uniform migration from the membrane into the catalyst layers. During high RH operation at OCV, in-plane cerium gradients arise due to variations in water content. These gradients may diminish the scavenging efficacy of cerium by reducing its proximity to generated radicals.

  18. Cerium migration during PEM fuel cell assembly and operation

    DOE PAGES

    Baker, Andrew M.; Torraco, Dennis; Judge, Elizabeth J.; ...

    2015-09-14

    Cerium migration between PEM fuel cell components is influenced by potential-driven mobility, ionic diffusion, and gradients in water content. These factors were investigated in ex situ experiments and in operating fuel cells. Potential-induced migration was measured ex situ in hydrated window cells. Cerium-containing MEAs were also fabricated and tested under ASTs. MEA disassembly and subsequent XRF analysis were used to observe rapid cerium migration during cell assembly and operation. During MEA hot pressing, humidification, and low RH operation at OCV, ionic diffusion causes uniform migration from the membrane into the catalyst layers. During high RH operation at OCV, in-plane ceriummore » gradients arise due to variations in water content. These gradients may diminish the scavenging efficacy of cerium by reducing its proximity to generated radicals.« less

  19. Fluid flow plate for decreased density of fuel cell assembly

    DOEpatents

    Vitale, Nicholas G.

    1999-01-01

    A fluid flow plate includes first and second outward faces. Each of the outward faces has a flow channel thereon for carrying respective fluid. At least one of the fluids serves as reactant fluid for a fuel cell of a fuel cell assembly. One or more pockets are formed between the first and second outward faces for decreasing density of the fluid flow plate. A given flow channel can include one or more end sections and an intermediate section. An interposed member can be positioned between the outward faces at an interface between an intermediate section, of one of the outward faces, and an end section, of that outward face. The interposed member can serve to isolate the reactant fluid from the opposing outward face. The intermediate section(s) of flow channel(s) on an outward face are preferably formed as a folded expanse.

  20. Fuel cell cooler assembly and edge seal means therefor

    DOEpatents

    Breault, Richard D.; Roethlein, Richard J.; Congdon, Joseph V.

    1980-01-01

    A cooler assembly for a stack of fuel cells comprises a fibrous, porous coolant tube holder sandwiched between and bonded to at least one of a pair of gas impervious graphite plates. The tubes are disposed in channels which pass through the holder. The channels are as deep as the holder thickness, which is substantially the same as the outer diameter of the tubes. Gas seals along the edges of the holder parallel to the direction of the channels are gas impervious graphite strips.

  1. High Energy Absorption Top Nozzle For A Nuclaer Fuel Assembly

    DOEpatents

    Sparrow, James A.; Aleshin, Yuriy; Slyeptsov, Aleksey

    2004-05-18

    A high energy absorption top nozzle for a nuclear fuel assembly that employs an elongated upper tubular housing and an elongated lower tubular housing slidable within the upper tubular housing. The upper and lower housings are biased away from each other by a plurality of longitudinally extending springs that are restrained by a longitudinally moveable piston whose upward travel is limited within the upper housing. The energy imparted to the nozzle by a control rod scram is mostly absorbed by the springs and the hydraulic affect of the piston within the nozzle.

  2. Gradient isolator for flow field of fuel cell assembly

    DOEpatents

    Ernst, William D.

    1999-01-01

    Isolator(s) include isolating material and optionally gasketing material strategically positioned within a fuel cell assembly. The isolating material is disposed between a solid electrolyte and a metal flow field plate. Reactant fluid carried by flow field plate channel(s) forms a generally transverse electrochemical gradient. The isolator(s) serve to isolate electrochemically a portion of the flow field plate, for example, transversely outward from the channel(s), from the electrochemical gradient. Further, the isolator(s) serve to protect a portion of the solid electrolyte from metallic ions.

  3. Gradient isolator for flow field of fuel cell assembly

    DOEpatents

    Ernst, W.D.

    1999-06-15

    Isolator(s) include isolating material and optionally gasketing material strategically positioned within a fuel cell assembly. The isolating material is disposed between a solid electrolyte and a metal flow field plate. Reactant fluid carried by flow field plate channel(s) forms a generally transverse electrochemical gradient. The isolator(s) serve to isolate electrochemically a portion of the flow field plate, for example, transversely outward from the channel(s), from the electrochemical gradient. Further, the isolator(s) serve to protect a portion of the solid electrolyte from metallic ions. 4 figs.

  4. Neutron collar calibration and evaluation for assay of LWR fuel assemblies containing burnable neutron absorbers

    SciTech Connect

    Henriksen, P.W.; Menlove, H.O.; Stewart, J.E.; Qiao, S.Z.; Wenz, T.R. ); Verrecchia, G.P.D. . Safeguards Directorate)

    1990-11-01

    The neutron coincidence collar is used to verify the uranium content in light water reactor fuel assemblies. An AmLi neutron source actively interrogates the fuel assembly to measure the {sup 235}U content and the {sup 238}U content can be verified from a passive neutron coincidence measurement. This report gives the collar calibration data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies both with and without cadmium liners. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and various fuel assembly sizes. The data were collected using the Los Alamos BWR and PWR test assemblies as well as fuel assemblies from several fuel fabrication facilities. 11 refs., 15 figs., 14 tabs.

  5. 96. SEED 1 FUEL ASSEMBLY FROM LOCATION L9 BEING REMOVED ...

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

    96. SEED 1 FUEL ASSEMBLY FROM LOCATION L-9 BEING REMOVED FROM REACTOR VESSEL BY MEANS OF FUEL EXTRACTION CRANE, JANUARY 7, 1960 - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

  6. Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

    DOEpatents

    Shore, Lawrence; Matlin, Ramail

    2009-12-22

    The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

  7. Total evaluation of in bundle void fraction measurement test of PWR fuel assembly

    SciTech Connect

    Hori, Keiichi; Miyazaki, Keiji; Akiyama, Yoshiei; Nishioka, Hiromasa; Takeda, Naoki

    1996-08-01

    Nuclear Power Engineering Corporation is performing the various proof or verification tests on safety and reliability of nuclear power plants under the sponsorship of the Ministry of International Trade and Industry. As one program of these Japanese national projects, an in bundle void fraction measurement test of a pressurized water reactor (PWR) fuel assembly was started in 1987 and finished at the end of 1994. The experiments were performed using the 5 x 5 square array rod bundle test sections. The rod bundle test section simulates the partial section and full length of a 17 x 17 type Japanese PWR fuel assembly. A distribution of subchannel averaged void fraction in a rod bundle test section was measured by the gamma-ray attenuation method using the stationary multi beam systems. The additional single channel test was performed to obtain the required information for the calibration of the rod bundle test data and the assessment of the void prediction method. Three test rod bundles were prepared to analyze an axial power distribution effect, an unheated rod effect, and a grid spacer effect. And, the obtained data were used for the assessment of the void prediction method relevant to the subchannel averaged void fraction of PWR fuel assemblies. This paper describes the outline of the experiments, the evaluation of the experimental data and the assessment of void prediction method.

  8. A CFD M&S PROCESS FOR FAST REACTOR FUEL ASSEMBLIES

    SciTech Connect

    Kurt D. Hamman; Ray A. Berry

    2008-09-01

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-e and SST (Menter) k-? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  9. Experimental characterization of aviation-fuel cavitation

    NASA Astrophysics Data System (ADS)

    Dunn, Patrick F.; Thomas, Flint O.; Davis, Michael P.; Dorofeeva, Irina E.

    2010-11-01

    The results of an experimental investigation of the gaseous cavitation of JP-8 aviation fuel in a converging-diverging nozzle are presented. Fuel cavitation is experimentally characterized by high-speed digital imaging, static pressure distributions, and nonintrusive void fraction and bubble velocity measurements. For comparative purposes, experiments were performed using distilled water and dodecane for the same nozzle and nozzle pressure ratios. Dodecane, the largest component of JP-8 by weight, served as its single-component surrogate. For each working fluid, the experiments examined two different flow regimes: an initially single-phase liquid flow in which no cavitation occurred and another that evolved into two-phase cavitating flow. Additional experiments were performed to study the effect of air bubbles injected into either water or JP-8 at the nozzle inlet. For a sufficiently low range of imposed back pressures, gaseous cavitation led to choked flow for each working fluid. The character of the cavitation in the three fluids was different. These differences are highlighted and plausible mechanisms responsible for the observed behavior are discussed.

  10. Nuclear fuel assemblies' deformations measurement by optoelectronic methods in cooling ponds

    NASA Astrophysics Data System (ADS)

    Senchenko, E. S.; Zavyalov, P. S.; Finogenov, L. V.; Khakimov, D. R.

    2013-12-01

    Increasing the reliability and life-time of nuclear fuel is actual problems for nuclear power engineering. It takes to provide the high geometric stability of nuclear fuel assemblies (FA) under exploitation, since various factors cause FA mechanical deformation (bending and twisting). To obtain the objective information and make recommendations for the FA design improvement one have to fulfill the post reactor FA analysis. Therefore it takes measurements of the FA geometric parameters in cooling ponds of nuclear power plants. As applied to this problem we have developed and investigated the different optoelectronic methods, namely, structured light method, television and shadow ones. In this paper effectiveness of these methods has been investigated using the special experimental test stand and fulfilled researches are described. The experimental results of FA measurements by different methods and recommendation for their usage is given.

  11. Stress and plastic deformation of MEA in fuel cells. Stresses generated during cell assembly

    NASA Astrophysics Data System (ADS)

    Bograchev, Daniil; Gueguen, Mikael; Grandidier, Jean-Claude; Martemianov, Serguei

    A linear elastic-plastic 2D model of fuel cell with hardening is developed for analysis of mechanical stresses in MEA arising in cell assembly procedure. The model includes the main components of real fuel cell (membrane, gas diffusion layers, graphite plates, and seal joints) and clamping elements (steel plates, bolts, nuts). The stress and plastic deformation in MEA are simulated with ABAQUS code taking into account the realistic clamping conditions. The stress distributions are obtained on the local and the global scales. The first one corresponds to the single tooth/channel structure. The global scale deals with features of the entire cell (the seal joint and the bolts). Experimental measurements of the residual membrane deformations have been provided at different bolts torques. The experimental data are in a good agreement with numerical predictions concerning the beginning of the plastic deformation.

  12. Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

    DOEpatents

    Bradley, John G.

    1982-01-01

    A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception.

  13. Cap assembly for a bundled tube fuel injector

    SciTech Connect

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-04-26

    A cap assembly for a bundled tube fuel injector includes an impingement plate and an aft plate that is disposed downstream from the impingement plate. The aft plate includes a forward side that is axially separated from an aft side. A tube passage extends through the impingement plate and the aft plate. A tube sleeve extends through the impingement plate within the tube passage towards the aft plate. The tube sleeve includes a flange at a forward end and an aft end that is axially separated from the forward end. A retention plate is positioned upstream from the impingement plate. A spring is disposed between the retention plate and the flange. The spring provides a force so as to maintain contact between at least a portion of the aft end of the tube sleeve and the forward side of the aft plate.

  14. Air Transport of Spent Nuclear Fuel (SNF) Assemblies

    SciTech Connect

    Haire, M.J.; Moses, S.D.; Shapovalov, V.I.; Morenko, A.

    2007-07-01

    Sometimes the only feasible means of shipping research reactor spent nuclear fuel (SNF) among countries is via air transport because of location or political conditions. The International Atomic Energy Agency (IAEA) has established a regulatory framework to certify air transport Type C casks. However, no such cask has been designed, built, tested, and certified. In lieu of an air transport cask, research reactor SNF has been transported using a Type B cask under an exemption with special arrangements for administrative and security controls. This work indicates that it may be feasible to transport commercial power reactor SNF assemblies via air, and that the cost is only about three times that of shipping it by railway. Optimization (i.e., reduction) of this cost factor has yet to be done. (authors)

  15. Forced-to-natural convection transition tests in parallel simulated liquid metal reactor fuel assemblies

    SciTech Connect

    Levin, A.E. ); Montgomery, B.H. )

    1990-01-01

    The Thermal-Hydraulic Out of Reactor Safety (THORS) Program at Oak Ridge National Laboratory (ORNL) had as its objective the testing of simulated, electrically heated liquid metal reactor (LMR) fuel assemblies in an engineering-scale, sodium loop. Between 1971 and 1985, the THORS Program operated 11 simulated fuel bundles in conditions covering a wide range of normal and off-normal conditions. The last test series in the Program, THORS-SHRS Assembly 1, employed two parallel, 19-pin, full-length, simulated fuel assemblies of a design consistent with the large LMR (Large Scale Prototype Breeder -- LSPB) under development at that time. These bundles were installed in the THORS Facility, allowing single- and parallel-bundle testing in thermal-hydraulic conditions up to and including sodium boiling and dryout. As the name SHRS (Shutdown Heat Removal System) implies, a major objective of the program was testing under conditions expected during low-power reactor operation, including low-flow forced convection, natural convection, and forced-to-natural convection transition at various powers. The THORS-SHRS Assembly 1 experimental program was divided up into four phases. Phase 1 included preliminary and shakedown tests, including the collection of baseline steady-state thermal-hydraulic data. Phase 2 comprised natural convection testing. Forced convection testing was conducted in Phase 3. The final phase of testing included forced-to-natural convection transition tests. Phases 1, 2, and 3 have been discussed in previous papers. The fourth phase is described in this paper. 3 refs., 2 figs.

  16. Performance of boiling water reactor fuel lead test assemblies to 35 MWd/kg U

    SciTech Connect

    Rowland, T.C.; Ikemoto, R.N.; Gehl, S.

    1986-01-01

    This joint Electric Power Research Institute/General Electric (EPRI/GE) fuel performance program involved thorough preirradiation characterization of fuel used in lead test assemblies (LTAs), detailed surveillance of their operation, and interim site examinations of the assemblies during reactor outages. The program originally included four GE-5 LTAs operating in the Peach Bottom-2 (PB-2) reactor. The program was later modified to include the pressurized fuel rod test assembly in the Peach Bottom-3 (PB-3) reactor. The program modification also included extending the operation of the PB-2 and PB-3 LTA fuel beyond normal discharge exposures. Results are summarized in the paper.

  17. Buoyancy-driven flow excursions in fuel assemblies

    SciTech Connect

    Laurinat, J.E.; Paul, P.K.; Menna, J.D.

    1995-12-31

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating moderator downward through channels in cylindrical fuel tubes. Powers were limited to prevent a flow excursion from occurring in one or more of these parallel channels. During full-power operation, limits prevented a boiling flow excursion from taking place. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increases beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of historical levels.

  18. Buoyancy-driven flow excursions in fuel assemblies

    SciTech Connect

    Laurinat, J.E.; Paul, P.K.; Menna, J.D.

    1995-09-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one of more of these parallel channels. During-full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increased beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations.

  19. Non-fuel assembly components: 10 CFR 61.55 classification for waste disposal

    SciTech Connect

    Migliore, R.J.; Reid, B.D.; Fadeff, S.K.; Pauley, K.A.; Jenquin, U.P.

    1994-09-01

    This document reports the results of laboratory radionuclide measurements on a representative group of non-fuel assembly (NFA) components for the purposes of waste classification. This document also provides a methodology to estimate the radionuclide inventory of NFA components, including those located outside the fueled region of a nuclear reactor. These radionuclide estimates can then be used to determine the waste classification of NFA components for which there are no physical measurements. Previously, few radionuclide inventory measurements had been performed on NFA components. For this project, recommended scaling factors were selected for the ORIGEN2 computer code that result in conservative estimates of radionuclide concentrations in NFA components. These scaling factors were based upon experimental data obtained from the following NFA components: (1) a pressurized water reactor (PWR) burnable poison rod assembly, (2) a PVM rod cluster control assembly, and (3) a boiling water reactor cruciform control rod blade. As a whole, these components were found to be within Class C limits. Laboratory radionuclide measurements for these components are provided in detail.

  20. 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.

  1. 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.

  2. Determination of the rod-wise fission gas release fraction in a complete fuel assembly using non-destructive gamma emission tomography

    NASA Astrophysics Data System (ADS)

    Holcombe, Scott; Andersson, Peter; Svärd, Staffan Jacobsson; Hallstadius, Lars

    2016-11-01

    A gamma tomography instrument has been developed at the Halden Boiling Water Reactor (HBWR) in cooperation between the Institute for Energy Technology, Westinghouse (Sweden) and Uppsala University. The instrument is used to record the gamma radiation field surrounding complete fuel assemblies and consists of a shielded enclosure with fixtures to accurately position the fuel and detector relative to each other. A High Purity Germanium detector is used for acquiring high-resolution spectroscopic data, allowing for analysis of multiple gamma-ray peaks. Using the data extracted from the selected peaks, tomographic reconstruction algorithms are used to reproduce the corresponding spatial gamma-ray source distributions within the fuel assembly. With this method, rod-wise data can be can be deduced without the need to dismantle the fuel. In this work, the tomographic device has been experimentally benchmarked for non-destructive rod-wise determination of the Fission Gas Release (FGR) fraction. Measurements were performed on the fuel-stack and gas-plenum regions of a complete fuel assembly, and quantitative tomographic reconstructions of the measurement data were performed in order to determine the rod-wise ratio of 85Kr in the gas plenum to 137Cs in the fuel stack. The rod-wise ratio of 85Kr/137Cs was, in turn, used to calculate the rod-wise FGR fraction. In connection to the tomographic measurements, the fuel rods were also measured individually using gamma scanning in order to provide an experimental benchmark for the tomographic method. Fuel rods from two donor driver fuel assemblies were placed into a nine-rod HBWR driver fuel assembly configuration. In order to provide a challenging measurement object and thus an appropriate benchmark for the tomographic method, five rods were taken from an assembly with a burnup of 51 MWd/kgUO2, and four rods were from an assembly with a burnup of 26 MWd/kgUO2. At the time of the measurements, the nine rods had cooled for

  3. Investigation on heavy liquid metal cooling of ADS fuel pin assemblies

    NASA Astrophysics Data System (ADS)

    Litfin, K.; Batta, A.; Class, A. G.; Wetzel, Th.; Stieglitz, R.

    2011-08-01

    In the framework of accelerator driven sub-critical reactor systems heavy liquid metals are considered as coolant for the reactor core and the spallation target. In particular lead or lead bismuth eutectic (LBE) exhibit efficient heat removal properties and high production rate of neutrons. However, the excellent heat conductivity of LBE-flows expressed by a low molecular Prandtl number of the order 10 -2 requires improved modeling of the turbulent heat transfer. Although various models for thermal hydraulics of LBE flows are existing, validated heat transfer correlations for ADS-relevant conditions are still missing. In order to validate the sub-channel codes and computational fluid dynamics codes used to design fuel assemblies, the comparison with experimental data is inevitable. Therefore, an experimental program composed of three major experiments, a single electrically heated rod, a 19-pin hexagonal water rod bundle and a LBE rod bundle, has been initiated at the Karlsruhe Liquid metal Laboratory (KALLA) of the Karlsruhe Institute of Technology, in order to quantify and separate the individual phenomena occurring in the momentum and energy transfer of a fuel assembly.

  4. Neutron collar calibration for assay of LWR (light-water reactor) fuel assemblies

    SciTech Connect

    Menlove, H.O.; Pieper, J.E.

    1987-03-01

    The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the /sup 235/U content, and the /sup 238/U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities.

  5. Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

    SciTech Connect

    Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir; Rael, Carlos D.; Desimone, David J.

    2014-09-18

    Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performed in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.

  6. Photon dose rates from spent fuel assemblies with relation to self- protection

    SciTech Connect

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

    1995-12-01

    Photon dose rates as a function of fission product decay times have been calculated for spent fuel assemblies typical of MTR-type research and test reactors. Based upon these dose rates, the length of time that a spent fuel assembly will be self-protecting (dose rate greater than 100 rem/h at 1 m in air) can be estimated knowing the mass of fuel burned, the fraction of fuel burned, and the fuel assembly specific power density. The calculated dose rates cover 20 years of fission product decay, spent fuel with up to 80% {sup 235}U burnup and assembly power densities ranging from 0.089 to 2.857 MW/kg{sup 235}U. Most of the results are unshielded dose rates at 1 m in air with some shielded dose rates at 40 cm in water. Dose rate sensitivity estimates have been evaluated for a variety of MTR fuel assembly designs and for uncertainties in both the physical and analytical models of the fuel assemblies.

  7. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    SciTech Connect

    Clarno, Kevin T; Hamilton, Steven P; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth; Pugmire, Dave; Dilts, Gary; Banfield, James E

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  8. Computational simulation of thermal hydraulic processes in the model LMFBR fuel assembly

    NASA Astrophysics Data System (ADS)

    Bayaskhalanov, M. V.; Merinov, I. G.; Korsun, A. S.; Vlasov, M. N.

    2017-01-01

    The aim of this study was to verify a developed software module on the experimental fuel assembly with partial blockage of the flow section. The developed software module for simulation of thermal hydraulic processes in liquid metal coolant is based on theory of anisotropic porous media with specially developed integral turbulence model for coefficients determination. The finite element method is used for numerical solution. Experimental data for hexahedral assembly with electrically heated smooth cylindrical rods cooled by liquid sodium are considered. The results of calculation obtained with developed software module for a case of corner blockade are presented. The calculated distribution of coolant velocities showed the presence of the vortex flow behind the blockade. Features vortex region are in a good quantitative and qualitative agreement with experimental data. This demonstrates the efficiency of the hydrodynamic unit for developed software module. But obtained radial coolant temperature profiles differ significantly from the experimental in the vortex flow region. The possible reasons for this discrepancy were analyzed.

  9. Fuel nozzle assembly for use as structural support for a duct structure in a combustor of a gas turbine engine

    DOEpatents

    Wiebe, David J; Fox, Timothy A

    2015-03-31

    A fuel nozzle assembly for use in a combustor apparatus of a gas turbine engine. An outer housing of the fuel nozzle assembly includes an inner volume and provides a direct structural connection between a duct structure and a fuel manifold. The duct structure defines a flow passage for combustion gases flowing within the combustor apparatus. The fuel manifold defines a fuel supply channel therein in fluid communication with a source of fuel. A fuel injector of the fuel nozzle assembly is provided in the inner volume of the outer housing and defines a fuel passage therein. The fuel passage is in fluid communication with the fuel supply channel of the fuel manifold for distributing the fuel from the fuel supply channel into the flow passage of the duct structure.

  10. Fuel sender assembly requiring no calibration and having reduced wear

    SciTech Connect

    Gaston, R.D.

    1990-05-15

    This patent describes a fuel sender assembly. It comprises: a float rod having a pivot portion and an arm portion; a housing member having a rod hole rotatably securing the pivot portion and having an arcuate slot radially disposed from the rod hole, the rod hole being sized to substantially prevent radial movement of the pivot portion; a resistance element secured to the housing member and having a first connection thereto; a carrier element received by the housing member, the carrier element having an interior portion aligned with the rod hole and receiving the pivot portion of the float rod, the carrier element having an exterior portion extending through the arcuate slot for receiving the arm portion, the exterior portion permitting limited movement of the arm portion relative to the housing member in a direction substantially parallel with the pivot portion; and spring contact means rigidly coupled to the interior portion of the carrier element for slidably contacting the resistance element as a second connection thereto.

  11. 3D laser inspection of fuel assembly grid spacers for nuclear reactors based on diffractive optical elements

    NASA Astrophysics Data System (ADS)

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

    2007-06-01

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

  12. Experimental Results for SimFuels

    SciTech Connect

    Buck, Edgar C.; Casella, Andrew M.; Skomurski, Frances N.; MacFarlan, Paul J.; Soderquist, Chuck Z.; Wittman, Richard S.; Mcnamara, Bruce K.

    2012-08-22

    Assessing the performance of Spent (or Used) Nuclear Fuel (UNF) in geological repository requires quantification of time-dependent phenomena that may influence its behavior on a time-scale up to millions of years. A high-level waste repository environment will be a dynamic redox system because of the time-dependent generation of radiolytic oxidants and reductants and the corrosion of Fe-bearing canister materials. One major difference between used fuel and natural analogues, including unirradiated UO2, is the intense radiolytic field. The radiation emitted by used fuel can produce radiolysis products in the presence of water vapor or a thin-film of water that may increase the waste form degradation rate and change radionuclide behavior. To study UNF, we have been working on producing synthetic UO2 ceramics, or SimFuels that can be used in testing and which will contain specific radionuclides or non-radioactive analogs so that we can test the impact of radiolysis on fuel corrosion without using actual spent fuel. Although, testing actual UNF would be ideal for understanding the long term behavior of UNF, it requires the use of hot cells and is extremely expensive. In this report, we discuss, factors influencing the preparation of SimFuels and the requirements for dopants to mimic the behavior of UNF. We have developed a reliable procedure for producing large grain UO2 at moderate temperatures. This process will be applied to a series of different formulations.

  13. Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

    NASA Technical Reports Server (NTRS)

    Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

    1971-01-01

    A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7 cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The experimental program consisted basically of measuring the differential neutron spectra and the changes in critical mass that accompanied the stepwise addition of (Li-7)3N, Hf, Ta, and W to a basic core fueled with U metal in a pin-type Ta honeycomb structure. In addition, experimental results were obtained on power distributions, control characteristics, neutron lifetime, and reactivity worths of numerous absorber, structural, and scattering materials.

  14. Method for providing concentricity of pilot fuel assembly in a combustor

    NASA Technical Reports Server (NTRS)

    Halila, Ely E. (Inventor); Anderson, Michael (Inventor); Martus, James A. (Inventor)

    2003-01-01

    Concentric installation of a pilot fuel assembly in an opening in a gas turbine combustor casing is achieved by providing a boss having at least two flat surfaces which are perpendicular to each other on the combustor casing surrounding the opening and a mounting flange having at least two flat surfaces which are perpendicular to each other on the pilot fuel assembly. The pilot fuel assembly is concentrically installed to the combustor casing by inserting the assembly into the combustor casing opening, and moving the pilot fuel assembly as far as it will go in a first direction substantially parallel to one of the flat boss surfaces. The distance between the other flat boss surface and one of the flat flange surfaces is then taken. Next, the pilot fuel assembly is moved in the direction opposite the first direction, at which point, the distance between the same two flat surfaces is again measured. Lastly, the pilot fuel assembly is located at a position where the distance between the two measuring surfaces is equal to the average of the first and second measurements. If desired, these steps can be repeated back and forth along an axis perpendicular to the first and second directions.

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

    SciTech Connect

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

    1996-05-01

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

  16. Use of Computational Fluid Dynamics (CFD) tools for fuel assembly analysis.

    SciTech Connect

    Garner, P. L.; Sofu, T.; Nuclear Engineering Division

    2004-01-01

    The STAR-CD computer program for Computational Fuel Dynamics (CFD) has been applied to the Russian pin-type fuel assemblies proposed as low enriched uranium (LEU) replacements for the high enriched uranium (HEU) (36%) IRT-3M fuel assemblies currently used in the WWR-SM reactor in Uzbekistan. For fuel assemblies containing twisted, finned pin-type fuel, STAR-CD was first used to model the single pin having the highest power density along with its associated coolant as an isolated unit cell. Velocity, pressure, temperature, heat flux, etc. were calculated on a detailed spatial basis in the coolant, cladding, and fuel. The model was then expanded to include multiple fuel pins; the computed motion of coolant from one portion of the assembly to another can reduce the peak temperatures below what one would compute using a single-pin model and, thus, change conclusions regarding the margin to onset of nucleate boiling. STAR-CD has also been applied to the IRT-3M tube-type fuel assemblies in the current HEU core.

  17. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    SciTech Connect

    G. W. Eccleston; H. O. Menlove; M. Abhold; M. Baker; J. Pecos

    1999-05-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency {sup 3}He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the {sup 240}Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual.

  18. Description and performance characteristics for the neutron Coincidence Collar for the verification of reactor fuel assemblies

    SciTech Connect

    Menlove, H.O.

    1981-08-01

    An active neutron interrogation method has been developed for the measurement of /sup 235/U content in fresh fuel assemblies. The neutron Coincidence Collar uses neutron interrogation with an AmLi neutron source and coincidence counting the induced fission reaction neutrons from the /sup 235/U. This manual describes the system components, operation, and performance characteristics. Applications of the Coincidence Collar to PWR and BWR types of reactor fuel assemblies are described.

  19. An analytical solution for the consideration of the effect of adjacent fuel assemblies; comparison of rectangular and hexagonal structures

    SciTech Connect

    Merk, B.; Rohde, U.

    2012-07-01

    A new analytical method is described to deal with the Leakage Environmental Effect. The method is based on the analytical solution of the two-group diffusion equation for two adjacent fuel assemblies. The quality of the results for this highly efficient method is demonstrated for square fuel assemblies. In additional tests the transferability of the concept to hexagonal VVER-440-type fuel assemblies is shown and a comparison between the results for rectangular and hexagonal assemblies is given. (authors)

  20. Fuel property effects on fuel/air mixing in an experimental diesel engine

    SciTech Connect

    Browne, K.R.; Patridge, I.M.; Greeves, G.

    1986-01-01

    Fuels of widely varying properties are studied by injection of a single and well defined spray into an experimental diesel engine. Three optical techniques were developed to visualise liquid fuel, fuel vapour, flame, soot and individual droplets and their associated vapour trails. Liquid core length measurements are presented for diesel fuel, toluene, ethanol and sunflower oil. Computer model predictions show that an increase of the fuel mid-boiling point by 40/sup 0/C gives a similar effect on liquid core length to an increase of 0.03mm in the nozzle hole diameter.

  1. Modeling depletion simulations for a high-burnup, highly heterogeneous BWR fuel assembly with scale

    SciTech Connect

    Smith, H. J.

    2012-07-01

    Extensive SCALE isotopic validation studies have been performed for various PWR fuel assembly designs and operating conditions, and to a lesser extent for BWR fuel assembly designs. However, no SCALE validation work has been documented for newer, highly heterogeneous BWR fuel assembly designs at high burnup. Isotopic benchmark calculations of the earlier, more geometrically uniform BWR fuel assemblies are less sensitive to simplification of the operating history details and certain modeling assumptions than heterogeneous fuel assemblies, particularly at high burnup. This analysis shows the capability of SCALE to simulate a complex highly heterogeneous SVEA96 Optima fuel assembly and illustrates the importance of the need for the highest possible accuracy and precision in isotope measurements intended to be used as benchmark-quality results. In addition, this analysis quantifies the impact of various modeling assumptions on the results. The sample for which the simulation results are reported here achieved a burnup 62 GWd/MTU and was analyzed as part of the MALIBU Extension program. (authors)

  2. Characterization of an Experimental Referee Broadened Specification (ERBS) aviation turbine fuel and ERBS fuel blends

    NASA Technical Reports Server (NTRS)

    Seng, G. T.

    1982-01-01

    Characterization data and comparisons of these data are presented for three individual lots of a research test fuel designated as an Experimental Referee Broadened Specification (ERBS) aviation turbine fuel. This research fuel, which is a blend of kerosene and hydrotreated catalytic gas oil, is a representation of a kerojet fuel with broadened properties. To lower the hydrogen content of the ERBS fuel, a blending stock, composed of xylene bottoms and hydrotreated catalytic gas oil, was developed and employed to produce two different ERBS fuel blends. The ERBS fuel blends and the blending stock were also characterized and the results for the blends are compared to those of the original ERBS fuel. The characterization results indicate that with the exception of the freezing point for ERBS lot 2, which was slightly high, the three lots, produced over a 2 year period, met all general fuel requirements. However, although the properties of the fuels were found to be fairly consistent, there were differences in composition. Similarly, all major requirements for the ERBS fuel blends were met or closely approached, and the properties of the blended fuels were found to generally reflect those expected for the proportions of ERBS fuel and blending stock used in their production.

  3. Rapid local adaptation mediates zooplankton community assembly in experimental mesocosms.

    PubMed

    Pantel, Jelena H; Duvivier, Cathy; Meester, Luc De

    2015-10-01

    Adaptive evolution can occur over similar timescales as ecological processes such as community assembly, but its particular effects on community assembly and structure and their magnitude are poorly understood. In experimental evolution trials, Daphnia magna were exposed to varying environments (presence and absence of fish and artificial macrophytes) for 2 months. Then, in a common gardening experiment, we compared zooplankton community composition when either experimentally adapted or D. magna from the original population were present. Local adaptation of D. magna significantly altered zooplankton community composition, leading to a suppression of abundances for some zooplankton taxa and facilitation for others. The effect size of D. magna adaptation was similar to that of adding fish or macrophytes to mesocosms, two important drivers of zooplankton community structure. Our results suggest that substantial amounts of variation in community composition in natural systems may be unexplained if evolutionary dynamics are ignored.

  4. INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP603) SHOWING CRANE ASSEMBLY ...

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

    INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP-603) SHOWING CRANE ASSEMBLY FOR TRANSFER PIT. INL PHOTO NUMBER NRTS-51-2404. Unknown Photographer, 5/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  5. Active Well Coincidence Counter measurements of enriched uranium fuel assemblies in scanning and stationary modes

    SciTech Connect

    Krick, M.S.; Cowder, L. ); Maltsev, V.; Chernikov, A.; Mokeenko, P.; D'yadkov, K.; Ivanov, V. Nuclear Power Plant, Zarechnyy ); Lagattu, A.; Lopatin, Y.; Czock, K.; Rundquist, D.; Pedraza, L. )

    1991-01-01

    Enriched uranium fuel assemblies were measured with an Active Well Coincidence Counter (AWCC) at the Beloyarskaya Nuclear Power Plant. Special AWCC inserts, electronics, and software were used. Stationary and scanning measurements were performed to establish calibrations and performance specifications for the assay of {sup 235}U and {sub 235}U/cm for BN600 fuel. 6 refs., 7 figs., 2 tabs.

  6. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    SciTech Connect

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  7. U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

    SciTech Connect

    Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.; Bowman, Stephen M.

    2016-09-01

    Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decay heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.

  8. Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

    SciTech Connect

    Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E.; Lovera, P.; Fleche, J. L.; Lacroix, M.; Carra, O.; Prele, G.

    2012-07-01

    Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO{sub 2} interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

  9. Prediction of dryout performance for boiling water reactor fuel assemblies based on subchannel analysis with the RINGS code

    SciTech Connect

    Knabe, P.; Wehle, F.

    1995-12-01

    A fuel assembly with a large critical power margin introduces flexibility into reload fuel management. Therefore, optimization of the bundle and spacer geometry to maximize the bundle critical power is an important design objective. With a view to reducing the extent of the complex full-scale tests usually carried out to determine the thermal-hydraulic characteristics of various assembly geometries, the subchannel analysis method was further developed with the Siemens RINGS code. The annular flow code predicts dryout power and dryout location by calculating the conditions at which the liquid film flow rate is reduced to zero, allowing for evaporation, droplet entrainment, and droplet deposition. Appropriate attention is paid to the modeling of spacer effects. Comparison with experimental data of 3 x 3 and 4 x 4 tests shows the capability of RINGS to predict the flow quality and mass flux in subchannels under typical boiling water reactor operating conditions. By using the RINGS code, experimental critical power data for 3 x 3, 4 x 4, 5 x 5, 7 x 7, 8 x 8, 9 x 9, and 10 x 10 fuel assemblies were successfully postcalculated.

  10. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    SciTech Connect

    Hamilton, Steven P; Clarno, Kevin T; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth

    2012-01-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms, such as neutron flux distribution, coolant conditions and assembly mechanical stresses, of traditional (single-pin) nuclear fuel performance simulation. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. With this novel capability, AMPFuel was used to model an entire 1717 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics). A full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 160 billion degrees of freedom for 10 loading steps. The single radiation transport calculation required about 50% of the time required to solve the thermo-mechanics with a single loading step, which demonstrates that it is feasible to incorporate, in a single code, a high-fidelity radiation transport capability with a high-fidelity nuclear fuel thermo-mechanics capability and anticipate acceptable computational requirements. The

  11. Method of preparing gas tags for identification of single and multiple failures of nuclear reactor fuel assemblies

    DOEpatents

    McCormick, Norman J.

    1976-01-01

    For use in the identification of failed fuel assemblies in a nuclear reactor, the ratios of the tag gas isotopic concentrations are located on curved surfaces to enable the ratios corresponding to failure of a single fuel assembly to be distinguished from those formed from any combination of two or more failed assemblies.

  12. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-12-31

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  13. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A. ); Blake, J.E.; Rush, G.C. )

    1990-01-01

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

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

    SciTech Connect

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  15. High mechanical performance of Areva upgraded fuel assemblies for PWR in USA

    SciTech Connect

    Gottuso, Dennis; Canat, Jean-Noel; Mollard, Pierre

    2007-07-01

    The merger of the product portfolios of the former Siemens and Framatome fuel businesses gave rise to a new family of PWR products which combine the best features of the different technologies to enhance the main performance of each of the existing products. In this way, the technology of each of the three main fuel assembly types usually delivered by AREVA NP, namely Mark-BW{sup TM}, HTP{sup TM} and AFA 3G{sup TM} has been enriched by one or several components from the others which contributes to improve their robustness and to enhance their performance. The combined experience of AREVA's products shows that the ROBUST FUELGUARD{sup TM}, the HMP{sup TM} end grid, the MONOBLOC{sup TM} guide tube, a welded structure, M5{sup R} material for every zirconium component and an upper QUICK-DISCONNECT{sup TM} are key features for boosting fuel assembly robustness. The ROBUST FUELGUARD benefits from a broad experience demonstrating its high efficiency in stopping debris. In addition, its mechanical strength has been enhanced and the proven blade design homogenizes the downstream flow distribution to strongly reduce excitation of fuel rods. The resistance to rod-to-grid fretting resistance of AREVA's new products is completed by the use of a lower HMP grid with 8 lines of contact to insure low wear. The Monobloc guide tube with a diameter maximized to strengthen the fuel assembly stiffness, excludes through its uniform outer geometry any local condition which could weaken guide tube straightness. The application of a welded cage to all fuel assemblies of the new family of products in combination with stiffer guide tubes and optimized hold-down assures each fuel assembly enhanced resistance to distortion. The combination of these features has been widely demonstrated as an effective method to reduce the risk of incomplete RCCA insertion and significantly reduce assembly distortion. Thanks to its enhanced performance, M5 alloy insures that all fuel assemblies in the family

  16. EASE (Experimental Assembly of Structures in EVA) overview of selected results

    NASA Technical Reports Server (NTRS)

    Akin, David L.

    1987-01-01

    Experimental Assembly of Structures in EVA (EASE) objectives, experimental protocol, neutral buoyancy simulation, task time distribution, assembly task performance, metabolic rate/biomedical readouts are summarized. This presentation is shown in charts, figures, and graphs.

  17. Experimental study of fuel sootiness effects on flashover.

    PubMed

    Tsai, Kuang-Chung; Chen, Hung-Hsiang

    2010-06-15

    Previous fire safety studies have demonstrated that flashover can result in severe injure and death and heat radiating back to a fuel is an important mechanism. Fuel sootiness dominates in radiative heat transfer. However, empirical correlations from previous investigations did not consider the fuel sootiness but nevertheless generated reasonably good predictions of flashover. In this study, a series of experiments was employed to examine fuel sootiness effects on flashover. The fuels used, in the order of their sootiness, were gasoline, n-hexane, iso-propanol and methanol. These fuels were filled in circular pans 100-320 mm in diameter to generate fires with different heat release rates and levels of sootiness. The pans were in 1/3 the size of the ISO 9705 test chamber. After ignition, the heat release rate (HRR), temperature inside the chamber, as well as heat flux on the floor and time to flashover (t(fo)) were determined. Experimental data show that HRR at flashover and t(fo) were strongly corrected and their relationship was independent of the fuel burned. Although heat feedback to the floor increased as fuel sootiness increased, consequently enhancing the burning of sooty fuels, flashover occurs only when the HRR at flashover criterion is reached.

  18. Investigation of a Shock Absorber for Safeguard of Fuel Assemblies Failure

    SciTech Connect

    Karalevicius, Renatas; Dundulis, Gintautas; Rimkevicius, Sigitas; Uspuras, Eugenijus

    2006-07-01

    The Ignalina NPP has two reactors. The Unit 1 was shut down, therefore the special equipment was designed for transportation of the fuel from Unit 1 to Unit 2. The fuel-loaded basket can drop during transportation. The special shock absorber was designed in order to avoid failure of fuel assemblies during transportation. In case of drop of fuel loaded basket, the failure of fuel assemblies can occur. This shock absorber was studied by scaled experiments at Lithuanian Energy Institute. Static and dynamic investigations of shock absorber are presented in this paper, including dependency of axial force versus axial compression. The finite element codes BRIGADE/Plus and ABAQUS/Explicit were used for analysis. Static simulation was used to optimize the dimensions of shock absorber. Dynamic analysis shows that shock absorber is capable to withstand the dynamic load for successful force suppression function in case of an accident. (authors)

  19. Experimental Test Plan for PWR Sister Rods in the High Burnup Spent Fuel Data Project

    SciTech Connect

    Montgomery, Rose; Scaglione, John M; Bevard, Bruce Balkcom; Hanson, Brady; Billone, Dr. Michael

    2016-01-01

    The High Burnup Spent Fuel Data project pulled 25 sister rods (9 from the project assemblies and 16 from similar HBU assemblies) for characterization. The 25 sister rods are all high burnup and cover the range of modern domestic cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of their long-term storage in a TN-32 dry storage cask (the cask rods ) to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup spent fuel.

  20. What happens inside a fuel cell? Developing an experimental functional map of fuel cell performance.

    PubMed

    Brett, Daniel J L; Kucernak, Anthony R; Aguiar, Patricia; Atkins, Stephen C; Brandon, Nigel P; Clague, Ralph; Cohen, Lesley F; Hinds, Gareth; Kalyvas, Christos; Offer, Gregory J; Ladewig, Bradley; Maher, Robert; Marquis, Andrew; Shearing, Paul; Vasileiadis, Nikos; Vesovic, Velisa

    2010-09-10

    Fuel cell performance is determined by the complex interplay of mass transport, energy transfer and electrochemical processes. The convolution of these processes leads to spatial heterogeneity in the way that fuel cells perform, particularly due to reactant consumption, water management and the design of fluid-flow plates. It is therefore unlikely that any bulk measurement made on a fuel cell will accurately represent performance at all parts of the cell. The ability to make spatially resolved measurements in a fuel cell provides one of the most useful ways in which to monitor and optimise performance. This Minireview explores a range of in situ techniques being used to study fuel cells and describes the use of novel experimental techniques that the authors have used to develop an 'experimental functional map' of fuel cell performance. These techniques include the mapping of current density, electrochemical impedance, electrolyte conductivity, contact resistance and CO poisoning distribution within working PEFCs, as well as mapping the flow of reactant in gas channels using laser Doppler anemometry (LDA). For the high-temperature solid oxide fuel cell (SOFC), temperature mapping, reference electrode placement and the use of Raman spectroscopy are described along with methods to map the microstructural features of electrodes. The combination of these techniques, applied across a range of fuel cell operating conditions, allows a unique picture of the internal workings of fuel cells to be obtained and have been used to validate both numerical and analytical models.

  1. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  2. The optimization of an AP1000 fuel assembly for the transmutation of plutonium and minor actinides

    NASA Astrophysics Data System (ADS)

    Washington, Jeremy A.

    The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. The goal of this thesis is to examine the potential of light water reactors for plutonium and minor actinides transmutation as a near-term solution. This thesis screens the available nuclear isotope database to identify potential absorbers as coatings on a transmutation fuel in a light water reactor. A spectral shift absorber coating tunes the neutron energy spectrum experienced by the underlying target fuel. Eleven different spectral shift absorbers (B4C, CdO, Dy2O3, Er 2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) have been selected for further evaluation. A model developed using the NEWT module of SCALE 6.1 code provided performance data for the burnup of the target fuel rods. Irradiation of the target fuels occurs in a Westinghouse 17x17 XL Robust Fuel Assembly over a 1400 Effective Full Power Days (EFPD) interval. The fuels evaluated in this thesis include PuO2, Pu3Si2, PuN, MOX, PuZrH, PuZrHTh, PuZrO 2, and PuUZrH. MOX (5 wt% PuO2), Pu0.31ZrH 1.6Th1.08, and PuZrO2MgO (8 wt%) are selected for detailed analysis in a multi-pin transmutation assembly. A coupled model optimized the resulting transmutation fuel elements. The optimization considered three stages of fuel assemblies containing target fuel pins. The first stage optimized four target fuel pins adjacent to the central instrumentation channel. The second stage evaluated a variety of assemblies with multiple target fuel pins and the third stage re-optimized target fuel pins in the second-stage assembly. A PuZrO2MgO (8 wt%) target fuel with a coating of Lu 2O3 resulted in the greatest reduction in curium-244

  3. Assaying Used Nuclear Fuel Assemblies Using Lead Slowing-Down Spectroscopy and Singular Value Decomposition

    SciTech Connect

    Kulisek, Jonathan A.; Anderson, Kevin K.; Casella, Andrew M.; Gesh, Christopher J.; Warren, Glen A.

    2013-04-01

    This study investigates the use of a Lead Slowing-Down Spectrometer (LSDS) for the direct and independent measurement of fissile isotopes in light-water nuclear reactor fuel assemblies. The current study applies MCNPX, a Monte Carlo radiation transport code, to simulate the measurement of the assay of the used nuclear fuel assemblies in the LSDS. An empirical model has been developed based on the calibration of the LSDS to responses generated from the simulated assay of six well-characterized fuel assemblies. The effects of self-shielding are taken into account by using empirical basis vectors calculated from the singular value decomposition (SVD) of a matrix containing the self-shielding functions from the assay of assemblies in the calibration set. The performance of the empirical algorithm was tested on version 1 of the Next-Generation Safeguards Initiative (NGSI) used fuel library consisting of 64 assemblies, as well as a set of 27 diversion assemblies, both of which were developed by Los Alamos National Laboratory. The potential for direct and independent assay of the sum of the masses of Pu-239 and Pu-241 to within 2%, on average, has been demonstrated.

  4. High Pressure Experimental Deformation in Talc Assemblies: Cheap, Easy, Wrong

    NASA Astrophysics Data System (ADS)

    Stewart, E. D.; Holyoke, C. W.; Kronenberg, A. K.; Newman, J.

    2011-12-01

    Early methods of applying high pressures in the Griggs piston-cylinder triaxial deformation apparatus employed solid confining media thought to be weak compared with the silicate samples under investigation. Griggs apparatus sample assemblies with talc as the confining medium have been used in large numbers of experimental studies of rock deformation at pressures of 1.0 to 2.0 GPa. Modern methods now use solid and molten salts as confining media, with flow strengths that are low for solid salts (at elevated temperatures) and zero for molten salts; yet, talc continues to be used when deformation microstructures are used to infer differential stress. Recent comparison experiments conducted in the Griggs apparatus and gas apparatus have yielded calibrations that enable accurate stress measurements using solid and molten salt assemblies. These calibrations demonstrate that differential stresses measured at high confining pressures using the Griggs apparatus are systematically high, yet they are correctable. They also indicate that a significant portion of the required stress correction is due to friction in the nested loading column of the apparatus. No other correction is required for molten salt assemblies and the offset in stress measurements using solid salt assemblies, due to the strength of solid salt, is considerably smaller than previously thought. Encouraged by these calibrations for solid and molten salt assemblies, we performed similar comparison experiments in the Griggs apparatus using traditional, easy-to-use talc assemblies with the goal of developing another calibration for high-pressure stress measurements. Following the same procedures as used in our earlier calibrations, we deformed molybdenum and TZM alloy cylinders using a talc assembly at the same temperatures and strain rates (600-1000{circ}C and 1{ast}10^{-4}/s) as used in gas apparatus experiments. The apparent strengths of the samples deformed below the talc dehydration temperature were at

  5. Partial discharge in a high voltage experimental test assembly

    SciTech Connect

    Koss, R.J.; Brainard, J.P.

    1998-07-01

    This study was initiated when a new type of breakdown occurred in a high voltage experimental test assembly. An anomalous current pulse was observed, which indicated partial discharges, some leading to total breakdowns. High voltage insulator defects are shown along with their effect on the electrostatic fields in the breakdown region. OPERA electromagnetic field modeling software is used to calculate the fields and present a cause for the discharge. Several design modifications are investigated and one of the simplest resulted in a 25% decrease in the field at the discharge surface.

  6. Preparation of a self-humidifying membrane electrode assembly for fuel cell and its performance analysis

    NASA Astrophysics Data System (ADS)

    Wang, Cheng; Mao, Zongqiang; Xu, Jingming; Xie, Xiaofeng; Yang, Lizhai

    2003-10-01

    A novel nano-porous material SiO2-gel was prepared. After being purified by H2O2, then protonized by H2SO4 and desiccated in vacuum, the SiO2-gel, mixed with Nafion solution, was coated between an electrode and a solid electrolyte, which made a new type of self-humidifying membrane electrode assembly. The SiO2 powder was characterized by FTIR, BET and XRD. The surface of the electrodes was characterized by SEM and EDS. The performances of the self-humidifying membrane electrodes were analyzed by polarization discharge and AC impedance under the operation modes of external humidification and self-humidification respectively. Experimental-results indicated that the SiO2 powder held super-hydrophilicity, and the layer of SiO2 and Nafion polymer between electrode and solid electrolyte expanded three-dimension electrochemistry reac-tion area, maintained stability of catalyst layer and enhanced back-diffusion of water from cathode to anode, so the PEM Fuel cell can generate electricity at self-humidification mode. The power density of single PEM fuel cell reached 1.5 W/cm2 under 0.2 Mpa, 70°C and dry hydrogen and oxygen.

  7. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    SciTech Connect

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  8. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

  9. Neutronic optimization in high conversion Th-{sup 233}U fuel assembly with simulated annealing

    SciTech Connect

    Kotlyar, D.; Shwageraus, E.

    2012-07-01

    This paper reports on fuel design optimization of a PWR operating in a self sustainable Th-{sup 233}U fuel cycle. Monte Carlo simulated annealing method was used in order to identify the fuel assembly configuration with the most attractive breeding performance. In previous studies, it was shown that breeding may be achieved by employing heterogeneous Seed-Blanket fuel geometry. The arrangement of seed and blanket pins within the assemblies may be determined by varying the designed parameters based on basic reactor physics phenomena which affect breeding. However, the amount of free parameters may still prove to be prohibitively large in order to systematically explore the design space for optimal solution. Therefore, the Monte Carlo annealing algorithm for neutronic optimization is applied in order to identify the most favorable design. The objective of simulated annealing optimization is to find a set of design parameters, which maximizes some given performance function (such as relative period of net breeding) under specified constraints (such as fuel cycle length). The first objective of the study was to demonstrate that the simulated annealing optimization algorithm will lead to the same fuel pins arrangement as was obtained in the previous studies which used only basic physics phenomena as guidance for optimization. In the second part of this work, the simulated annealing method was used to optimize fuel pins arrangement in much larger fuel assembly, where the basic physics intuition does not yield clearly optimal configuration. The simulated annealing method was found to be very efficient in selecting the optimal design in both cases. In the future, this method will be used for optimization of fuel assembly design with larger number of free parameters in order to determine the most favorable trade-off between the breeding performance and core average power density. (authors)

  10. An experimental evaluation of a small fusion fuel cleanup system

    SciTech Connect

    Holtslander, W.J.; Johnson, R.E.; Gravelle, F.B.; Schultz, C.M.

    1986-01-01

    Small tritium-burning experimental tokamaks will require some means of handling and purifying the deuterium-tritium fuel. A simple purification system would allow reinjection of fuel, minimize tritium inventory on site, and reduce the number of shipments of tritium to and from the tokamak site. This could simplify the licensing and safety aspects for sites unsuited to large inventories of tritium. At the request of the Canadian Fusion Fuels Technology Project, a number of conceptual designs of fusion fuel cleanup systems were prepared. These designs were based on handling 5000-Ci batches of fuel containing helium (2%), water (0.4%), oxygen and nitrogen (0.1% each), and carbon oxides and methane (0.5% each). The purified fuel was to have impurity concentrations no greater than 1% helium and 0.1% total for the remainder. Six conceptual designs were prepared and evaluated. In each of these, the fuel from the tokamak was diluted to {approximately}25% in helium prior to processing. The basis of the purification cycle was to dilute the fuel with helium as a carrier gas, remove all of the hydrogen and impurities, and regenerate pure fuel for reuse. The preferred design consisted of a gas circulation loop comprising an expansion tank, a pump, and a number of purification units, a uranium bed, a zirconium-aluminum getter bed, and two catalyst beds, Pt/Pd and CuO/MnO{sub 2}. This paper summarizes an experimental evaluation of this system using hydrogen and nontriated impurities. 1 ref.

  11. Integral gas seal for fuel cell gas distribution assemblies and method of fabrication

    DOEpatents

    Dettling, Charles J.; Terry, Peter L.

    1985-03-19

    A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

  12. Method of fabricating an integral gas seal for fuel cell gas distribution assemblies

    DOEpatents

    Dettling, Charles J.; Terry, Peter L.

    1988-03-22

    A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

  13. An experimental electrical generating unit using sugarcane bagasse as fuel

    SciTech Connect

    Elkoury, J.M.

    1980-12-01

    The purpose of this paper is to present the alternatives that exist within the Puerto Rico Electric Power Authority to develop an experimental electrical generating unit which would use sugarcane bagasse as fuel. The study includes a comparison between the sugarcane bagasse and other fuels, the location of an experimental electrical generating unit with respect to the sugarcane fields, the transportation of the bagasse and the generating equipment available for this project in terms of its fisical condition. This latter part would include any modifications in the equipment which we would have to undertake in order to carry out the study.

  14. Microstructural Analysis of an HT9 Fuel Assembly Duct Irradiated in FFTF to 155 Dpa at 443ºC

    SciTech Connect

    Bulent H. Sencer; James I Cole; John R. Kennedy; Stuart A. Maloy; Frank A. Garner

    2009-09-01

    The majority of published data on the irradiation response of ferritic/martensitic steels has been derived from simple free-standing specimens irradiated in experimental assemblies under well-defined and near-constant conditions, while components of long-lived fuel assemblies are more complex in shape and will experience progressive changes in environmental conditions. To insure that the resistance of HT9 to void swelling is maintained under more realistic operating conditions, this study addresses the radiation-induced microstructure of an HT9 ferritic/martensitic (F/M) steel hexagon duct that was examined following a six-year irradiation campaign of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF). The calculated irradiation exposure and operating temperature of the duct location examined were ~155 dpa at ~443ºC. It was found that dislocation networks were contained predominantly a/2<111> Burgers vector. Surprisingly, for such a large irradiation dose, type a<100> interstitial loops were observed at relatively high density. Additionally, a high density of precipitation was observed. These two microstructural characteristics may have contributed to the rather low swelling level of 0.3%. It appears that the inherent swelling resistance of this alloy observed in specimens irradiated under non-varying experimental conditions is not significantly degraded compared to time-dependent variations in neutron flux-spectra, temperature and stress state that are characteristic of actual reactor components.

  15. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    SciTech Connect

    Chodak, III, Paul

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  16. Heat transfer analysis of fuel assemblies in a heterogeneous gas core nuclear rocket

    NASA Technical Reports Server (NTRS)

    Watanabe, Yoichi; Appelbaum, Jacob; Diaz, Nils; Maya, Isaac

    1991-01-01

    Heat transfer problems of a heterogeneous gaseous core nuclear rocket were studied. The reactor core consists of 1.5-m long hexagonal fuel assemblies filled with pressurized uranium tetrafluoride (UF4) gas. The fuel gas temperature ranges from 3500 to 7000 K at a nominal operating condition of 40 atm. Each fuel assembly has seven coolant tubes, through which hydrogen propellant flows. The propellant temperature is not constrained by the fuel temperature but by the maximum temperature of the graphite coolant tube. For a core achieving a fission power density of 1000 MW/cu m, the propellant core exit temperature can be as high as 3200 K. The physical size of a 1250 MW gaseous core nuclear rocket is comparable with that of a NERVA-type solid core nuclear rocket. The engine can deliver a specific impulse of 1020 seconds and a thrust of 330 kN.

  17. Morphological features (defects) in fuel cell membrane electrode assemblies

    NASA Astrophysics Data System (ADS)

    Kundu, S.; Fowler, M. W.; Simon, L. C.; Grot, S.

    Reliability and durability issues in fuel cells are becoming more important as the technology and the industry matures. Although research in this area has increased, systematic failure analysis, such as a failure modes and effects analysis (FMEA), are very limited in the literature. This paper presents a categorization scheme of causes, modes, and effects related to fuel cell degradation and failure, with particular focus on the role of component quality, that can be used in FMEAs for polymer electrolyte membrane (PEM) fuel cells. The work also identifies component defects imparted on catalyst-coated membranes (CCM) by manufacturing and proposes mechanisms by which they can influence overall degradation and reliability. Six major defects have been identified on fresh CCM materials, i.e., cracks, orientation, delamination, electrolyte clusters, platinum clusters, and thickness variations.

  18. Process for recycling components of a PEM fuel cell membrane electrode assembly

    DOEpatents

    Shore, Lawrence [Edison, NJ

    2012-02-28

    The membrane electrode assembly (MEA) of a PEM fuel cell can be recycled by contacting the MEA with a lower alkyl alcohol solvent which separates the membrane from the anode and cathode layers of the assembly. The resulting solution containing both the polymer membrane and supported noble metal catalysts can be heated under mild conditions to disperse the polymer membrane as particles and the supported noble metal catalysts and polymer membrane particles separated by known filtration means.

  19. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    SciTech Connect

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.; Bounds, J.A.; Kimpland, R.H.; Damjanovich, R.P.; Jaegers, P.J.

    1997-08-01

    Experiments were performed to measure a variety of parameters for SHEBA: behavior of the facility during transient and steady-state operation; characteristics of the SHEBA fuel; delayed-critical solution height vs solution temperature; initial reactor period and reactivity vs solution height; calibration of power level vs reactor power instrumentation readings; flux profile in SHEBA; radiation levels and neutron spectra outside the assembly for code verification and criticality alarm and dosimetry purposes; and effect on reactivity of voids in the fuel.

  20. Fabrication Method for Laboratory-Scale High-Performance Membrane Electrode Assemblies for Fuel Cells.

    PubMed

    Sassin, Megan B; Garsany, Yannick; Gould, Benjamin D; Swider-Lyons, Karen E

    2017-01-03

    Custom catalyst-coated membranes (CCMs) and membrane electrode assemblies (MEAs) are necessary for the evaluation of advanced electrocatalysts, gas diffusion media (GDM), ionomers, polymer electrolyte membranes (PEMs), and electrode structures designed for use in next-generation fuel cells, electrolyzers, or flow batteries. This Feature provides a reliable and reproducible fabrication protocol for laboratory scale (10 cm(2)) fuel cells based on ultrasonic spray deposition of a standard Pt/carbon electrocatalyst directly onto a perfluorosulfonic acid PEM.

  1. Feasibility study on the verification of fresh fuel assemblies in shipping containers

    SciTech Connect

    Swinth, K.L.; Tanner, J.E.

    1990-09-01

    The purpose of this study was to examine the feasibility of using various nondestructive measurement techniques to determine the presence of fuel assemblies inside shipping containers and to examine the feasibility of measuring the fissile content of the containers. Passive and active techniques based on both gamma and neutron assay were examined. In addition, some experiments and calculations were performed to evaluate neutron techniques. Passive counting of the 186 keV gamma from {sup 235}U is recommended for use as an attributes measurement technique. Experiments and studies indicated that a bismuth germanate (BGO) scintillator is the preferred detector. A properly designed system based on this detector will provide a compact detector that can selectively verify fuel assemblies within a shipping container while the container is in a stack of similarly loaded containers. Missing fuel assemblies will be readily detected, but gamma counting of assemblies cannot detect changes in the fissile content of the inner rods in an assembly. If a variables technique is required, it is recommended that more extensive calculations be performed and removal of the outer shipping container be considered. Marking (sealing) of the assemblies with a uniquely identifiable transponder was also considered. This would require the development of procedures that would assure proper application and removal of the seal. When change to a metal outer container occurs, the technique will no longer be useful unless a radiolucent window is included in the container. 20 refs., 7 figs., 2 tabs.

  2. Facile and green assembly of nanocomposite membranes for fuel cells.

    PubMed

    Quartarone, Eliana; Villa, Davide Carlo; Angioni, Simone; Mustarelli, Piercarlo

    2015-02-04

    We report on a facile spray deposition method, which allows obtaining nanocomposite membranes for high-temperature polymer fuel cells characterized by high homogeneity and excellent proton conductivity. The proposed method is also green, as it requires much smaller amounts of solvents with respect to standard casting.

  3. Fuel cell system including a unit for electrical isolation of a fuel cell stack from a manifold assembly and method therefor

    DOEpatents

    Kelley; Dana A. , Farooque; Mohammad , Davis; Keith

    2007-10-02

    A fuel cell system with improved electrical isolation having a fuel cell stack with a positive potential end and a negative potential, a manifold for use in coupling gases to and from a face of the fuel cell stack, an electrical isolating assembly for electrically isolating the manifold from the stack, and a unit for adjusting an electrical potential of the manifold such as to impede the flow of electrolyte from the stack across the isolating assembly.

  4. Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

    SciTech Connect

    Ham, Y.S.; Kerr, P.; Sitaraman, S.; Swan, R.; Rossa, R.; Liljenfeldt, H.

    2015-07-01

    The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)

  5. An experimental/analytical comparison of strains in encapsulated assemblies

    SciTech Connect

    Guess, T.R.; Burchett, S.N.

    1991-11-01

    A combined experimental and analytical study of strains developed in encapsulated assemblies during casting, curing and thermal excursions is described. The experimental setup, designed to measure in situ strains, consisted of thin, closed-end, Kovar tubes that were instrumented with strain gages and thermocouples before being over-cast with a polymeric encapsulant. Four bisphenol A (three diethanolamine cured and one anhydride cured) epoxy-based materials and one urethane elastomeric material were studied. After cure of the encapsulant, tube strains were measured over the temperature range of {minus}55{degrees}C to 90{degrees}C. The thermal excursion experiments were then numerically modeled using finite element analyses and the computed strains were compared to the experimental strains. The predicted strains were over estimated (conservative) when a linear, elastic, temperature-dependent material model was assumed for the encapsulant and the stress free temperature T{sub i} was assumed to correspond to the cure temperature {Tc} of the encapsulant. Very good agreement was obtained with linear elastic calculations provided that the stress free temperature corresponded to the onset of the glassy-to-rubbery transition range of the encapsulant. Finally, excellent agreement was obtained in one of the materials (828/DEA) when a viscoelastic material model was utilized and a stress free temperature corresponding to the cure temperature was assumed. 13 refs., 20 figs., 3 tabs.

  6. Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

    NASA Astrophysics Data System (ADS)

    Burdo, James S.

    This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the

  7. Automated closure system for nuclear reactor fuel assemblies

    DOEpatents

    Christiansen, David W.; Brown, William F.

    1985-01-01

    A welder for automated closure of fuel pins by a pulsed magnetic process in which the open end of a length of cladding is positioned within a complementary tube surrounded by a pulsed magnetic welder. Seals are provided at each end of the tube, which can be evacuated or can receive tag gas for direct introduction to the cladding interior. Loading of magnetic rings and end caps is accomplished automatically in conjunction with the welding steps carried out within the tube.

  8. Fuel nozzle assembly for use in turbine engines and methods of assembling same

    DOEpatents

    Uhm, Jong Ho; Johnson, Thomas Edward

    2015-02-03

    A fuel nozzle for use with a turbine engine is described herein. The fuel nozzle includes a housing that is coupled to a combustor liner defining a combustion chamber. The housing includes an endwall that at least partially defines the combustion chamber. A plurality of mixing tubes extends through the housing for channeling fuel to the combustion chamber. Each mixing tube of the plurality of mixing tubes includes an inner surface that extends between an inlet portion and an outlet portion. The outlet portion is oriented adjacent the housing endwall. At least one of the plurality of mixing tubes includes a plurality of projections that extend outwardly from the outlet portion. Adjacent projections are spaced a circumferential distance apart such that a groove is defined between each pair of circumferentially-apart projections to facilitate enhanced mixing of fuel in the combustion chamber.

  9. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    SciTech Connect

    Margaret A. Marshall

    2012-09-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  10. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    SciTech Connect

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  11. PBF Reactor Building (PER620). Detail of fuel test assembly in ...

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

    PBF Reactor Building (PER-620). Detail of fuel test assembly in preparation for test. When complete, it will fit into in-pile tube. The maximum outside diameter of which must be about 8.25 inches. Date: 1982. INEEL negative no. 82-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  12. Measuring the Multiplication of Spent Fuel Assemblies – It’s easier than you think!

    SciTech Connect

    Tobin, Stephen Joseph

    2016-11-09

    This is a set of eight slides which advertise how easy it can be to measure the multiplication of a spent fuel assembly. A robust (fission chambers), rapid (under 15 minutes), direct (multiplication is measured, not photons from fission fragments) measurement of multiplication is possible.

  13. Effect of assembly error of bipolar plate on the contact pressure distribution and stress failure of membrane electrode assembly in proton exchange membrane fuel cell

    NASA Astrophysics Data System (ADS)

    Liu, Dong'an; Peng, Linfa; Lai, Xinmin

    In practice, the assembly error of the bipolar plate (BPP) in a PEM fuel cell stack is unavoidable based on the current assembly process. However its effect on the performance of the PEM fuel cell stack is not reported yet. In this study, a methodology based on FEA model, "least squares-support vector machine (LS-SVM)" simulation and statistical analysis is developed to investigate the effect of the assembly error of the BPP on the pressure distribution and stress failure of membrane electrode assembly (MEA). At first, a parameterized FEA model of a metallic BPP/MEA assembly is established. Then, the LS-SVM simulation process is conducted based on the FEA model, and datasets for the pressure distribution and Von Mises stress of MEA are obtained, respectively for each assembly error. At last, the effect of the assembly error is obtained by applying the statistical analysis to the LS-SVM results. A regression equation between the stress failure and the assembly error is also built, and the allowed maximum assembly error is calculated based on the equation. The methodology in this study is beneficial to understand the mechanism of the assembly error and can be applied to guide the assembly process for the PEM fuel cell stack.

  14. PEM fuel cell cost minimization using ``Design For Manufacture and Assembly`` techniques

    SciTech Connect

    Lomax, F.D. Jr.; James, B.D.; Mooradian, R.P.

    1997-12-31

    Polymer Electrolyte Membrane (PEM) fuel cells fueled with direct hydrogen have demonstrated substantial technical potential to replace Internal Combustion Engines (ICE`s) in light duty vehicles. Such a transition to a hydrogen economy offers the potential of substantial benefits from reduced criteria and greenhouse emissions as well as reduced foreign fuel dependence. Research conducted for the Ford Motor Co. under a US Department of Energy contract suggests that hydrogen fuel, when used in a fuel cell vehicle (FCV), can achieve a cost per vehicle mile less than or equal to the gasoline cost per mile when used in an ICE vehicle. However, fuel cost parity is not sufficient to ensure overall economic success: the PEM fuel cell power system itself must be of comparable cost to the ICE. To ascertain if low cost production of PEM fuel cells is feasible, a powerful set of mechanical engineering tools collectively referred to as Design for Manufacture and Assembly (DFMA) has been applied to several representative PEM fuel cell designs. The preliminary results of this work are encouraging, as presented.

  15. Fuel-rich, catalytic reaction experimental results. [fuel development for high-speed civil transport aircraft

    NASA Technical Reports Server (NTRS)

    Rollbuhler, Jim

    1991-01-01

    Future aeropropulsion gas turbine combustion requirements call for operating at very high inlet temperatures, pressures, and large temperature rises. At the same time, the combustion process is to have minimum pollution effects on the environment. Aircraft gas turbine engines utilize liquid hydrocarbon fuels which are difficult to uniformly atomize and mix with combustion air. An approach for minimizing fuel related problems is to transform the liquid fuel into gaseous form prior to the completion of the combustion process. Experimentally obtained results are presented for vaporizing and partially oxidizing a liquid hydrocarbon fuel into burnable gaseous components. The presented experimental data show that 1200 to 1300 K reaction product gas, rich in hydrogen, carbon monoxide, and light-end hydrocarbons, is formed when flowing 0.3 to 0.6 fuel to air mixes through a catalyst reactor. The reaction temperatures are kept low enough that nitrogen oxides and carbon particles (soot) do not form. Results are reported for tests using different catalyst types and configurations, mass flowrates, input temperatures, and fuel to air ratios.

  16. PWR internal flow modeling with fuel assemblies details

    SciTech Connect

    Popov, E.; Yan, J.; Karoutas, Z.; Gehin, J.; Brewster, R.; Baglietto, E.

    2012-07-01

    This study is an example of a massive parallel computing of the coolant flow in a nuclear reactor. It resolves the flow velocities in each assembly on pin level and predicts the flow distribution in complex geometries such as the lower and upper reactor plenums. The size of the developed model (1.035 billion cells) required the runs to be executed on the NCCS clusters (www.nccs.gov). STAR-CCM+ code (www.ed-adapco.com) was installed on two clusters: JAGUARXT5 and FROST, both of which were capable of executing this model. (authors)

  17. Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

    SciTech Connect

    McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard; Grey, Carissa; Engelhardt, Charles; Saltzstein, Sylvia J.; Sorenson, Ken B.

    2016-05-01

    This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

  18. Welding fixture for nuclear fuel pin cladding assemblies

    DOEpatents

    Oakley, D.J.; Feld, S.H.

    1984-02-22

    A welding fixture is described for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

  19. Welding fixture for nuclear fuel pin cladding assemblies

    DOEpatents

    Oakley, David J.; Feld, Sam H.

    1986-01-01

    A welding fixture for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

  20. Numerical and Experimental Model Studies on Thermal Hydraulic Behavior of FBR Internal Core Catcher Assembly

    SciTech Connect

    Sanjay Kumar Das; Anil Kumar Sharma; Jasmin Sudha, A.; Punitha, G.; Lydia, G.; Somayajulu, P.A.; Murthy, S.S.; Malarvizhi, B.; Gopalakrishnan, V.; Harvey, J.; Kasinathan, N.; Rajan, M.

    2006-07-01

    Core Catcher is provided as an in-vessel core debris retention device to collect, support, cool and maintain in sub-critical configuration, the generated core debris from fuel melting due to certain postulated Beyond Design Basis Events (BDBE) for Fast Breeder Reactor (FBR). This also acts as a barrier to prevent settling of debris on main vessel and keeps its maximum temperature within acceptable creep range. Heat transfer by natural convection in the core catcher assembly has been assessed numerically and through water experiments using geometrically similar configuration. Resistive heating elements are used in experiment as heat source to simulate debris decay heat on core catcher. Series of experiments were carried out for two configurations referred as geometry A and geometry B. The later configuration showed enhanced natural convective heat transfer from the lower plenum of the vessel. Temperatures were monitored at critical positions and compared with numerical evaluation. Numerically evaluated flow fields and isotherms are compared with experimental data for specific steady state temperatures on heat source plate. Numerical results are found to be in good agreement with that obtained from experiments. The combined efforts of numerical and experimental work conclude core catcher assembly with geometry B to be more suitable. (authors)

  1. Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

    NASA Technical Reports Server (NTRS)

    Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

    1972-01-01

    A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The range of the previous experimental investigations has been expanded to include the reactivity effects of:(1) surrounding the reactor with 15.24 cm (6 in.) of polyethylene, (2) reducing the heights of a portion of the upper and lower axial reflectors by factors of 2 and 4, (3) adding 45 kg of W to the core uniformly in two steps, (4) adding 9.54 kg of Ta to the core uniformly, and (5) inserting 2.3 kg of polyethylene into the core proper and determining the effect of a Ta addition on the polyethylene worth.

  2. Verifying nuclear fuel assemblies in wet storages on a partial defect level: A software simulation tool for evaluating the capabilities of the Digital Cherenkov Viewing Device

    NASA Astrophysics Data System (ADS)

    Grape, Sophie; Jacobsson Svärd, Staffan; Lindberg, Bo

    2013-01-01

    The Digital Cherenkov Viewing Device (DCVD) is an instrument that records the Cherenkov light emitted from irradiated nuclear fuels in wet storages. The presence, intensity and pattern of the Cherenkov light can be used by the International Atomic Energy Agency (IAEA) inspectors to verify that the fuel properties comply with declarations. The DCVD is since several years approved by the IAEA for gross defect verification, i.e. to control whether an item in a storage pool is a nuclear fuel assembly or a non-fuel item [1]. Recently, it has also been endorsed as a tool for partial defect verification, i.e. to identify if a fraction of the fuel rods in an assembly have been removed or replaced. The latter recognition was based on investigations of experimental studies on authentic fuel assemblies and of simulation studies on hypothetic cases of partial defects [2]. This paper describes the simulation methodology and software which was used in the partial defect capability evaluations. The developed simulation procedure uses three stand-alone software packages: the ORIGEN-ARP code [3] used to obtain the gamma-ray spectrum from the fission products in the fuel, the Monte Carlo toolkit Geant4 [4] for simulating the gamma-ray transport in and around the fuel and the emission of Cherenkov light, and the ray-tracing programme Zemax [5] used to model the light transport through the assembly geometry to the DCVD and to mimic the behaviour of its lens system. Furthermore, the software allows for detailed information from the plant operator on power and/or burnup distributions to be taken into account to enhance the authenticity of the simulated images. To demonstrate the results of the combined software packages, simulated and measured DCVD images are presented. A short discussion on the usefulness of the simulation tool is also included.

  3. Experimental insight into the process of parasite community assembly

    Technology Transfer Automated Retrieval System (TEKTRAN)

    1. Community assembly is a fundamental process that has long been a central focus in ecology. Extending community assembly theory to communities of co-infecting parasites, we used a gastrointestinal nematode removal experiment in free-ranging African buffalo to examine community assembly patterns an...

  4. Fuel moisture influences on fire-altered carbon in masticated fuels: An experimental study

    NASA Astrophysics Data System (ADS)

    Brewer, Nolan W.; Smith, Alistair M. S.; Hatten, Jeffery A.; Higuera, Philip E.; Hudak, Andrew T.; Ottmar, Roger D.; Tinkham, Wade T.

    2013-03-01

    Biomass burning is a significant contributor to atmospheric carbon emissions but may also provide an avenue in which fire-affected ecosystems can accumulate carbon over time, through the generation of highly resistant fire-altered carbon. Identifying how fuel moisture, and subsequent changes in the fire behavior, relates to the production of fire-altered carbon is important in determining how persistent charred residues are following a fire within specific fuel types. Additionally, understanding how mastication (mechanical forest thinning) and fire convert biomass to black carbon is essential for understanding how this management technique, employed in many fire-prone forest types, may influence stand-level black carbon in soils. In this experimental study, 15 masticated fuel beds, conditioned to three fuel moisture ranges, were burned, and production rates of pyrogenic carbon and soot-based black carbon were evaluated. Pyrogenic carbon was determined through elemental analysis of the post-fire residues, and soot-based black carbon was quantified with thermochemical methods. Pyrogenic carbon production rates ranged from 7.23% to 8.67% relative to pre-fire organic carbon content. Black carbon production rates averaged 0.02% in the 4-8% fuel moisture group and 0.05% in the 13-18% moisture group. A comparison of the ratio of black carbon to pyrogenic carbon indicates that burning with fuels ranging from 13% to 15% moisture content resulted in a higher proportion of black carbon produced, suggesting that the precursors to black carbon were indiscriminately consumed at lower fuel moistures. This research highlights the importance of fuel moisture and its role in dictating both the quantity and quality of the carbon produced in masticated fuel beds.

  5. Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

    NASA Astrophysics Data System (ADS)

    Artnak, Edward Joseph, III

    This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

  6. Acceptance of non-fuel assembly hardware by the Federal Waste Management System

    SciTech Connect

    Not Available

    1990-03-01

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high-level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. 14 refs., 12 figs., 43 tabs.

  7. CFD prediction of flow and phase distribution in fuel assemblies with spacers

    SciTech Connect

    Anglart, H.; Nylund, O.; Kurul, N.

    1995-09-01

    This paper is concerned with the modeling and computation of multi-dimensional two-phase flows in BWR fuel assemblies. The modeling principles are presented based on using a two-fluid model in which lateral interfacial effects are accounted for. This model has been used to evaluate the velocity fields of both vapor and liquid phases, as well as phase distribution, between fuel elements in geometries similar to BWR fuel bundles. Furthermore, this model has been used to predict, in a detailed mechanistic manner, the effects of spacers on flow and phase distribution between, and pressure drop along, fuel elements. The related numerical simulations have been performed using a CFD computer code, CFDS-FLOW3D.

  8. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    SciTech Connect

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio; Davila, Jesus

    2015-07-23

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides {sup 232}Th, {sup 238}U and {sup 237}Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  9. Differential die-away technique for determination of the fissile contents in spent fuel assembly

    SciTech Connect

    Lee, Tachoon; Menlove, Howard O; Swinhoe, Nartyn T; Tobin, Stephen J

    2010-01-01

    Monte Carlo simulations were performed for the differential die-away (DDA) technique to quantify its capability to measure the fissile contents in spent fuel assemblies of 64 different cases in terms of initial enrichment, burnup, and cooling time. The DDA count rate varies according to the contents of fissile isotopes such as {sup 235}U, {sup 239}Pu, and {sup 241}Pu contained in the spent fuel assembly. The effective {sup 239}Pu concept was introduced to quantify the total fissile mass of spent fuel by weighting the relative signal contributions of {sup 235}U and {sup 241}Pu compared to that of {sup 239}Pu. The Monte Carlo simulation results show that the count rate of the DDA instrument for a spent fuel assembly of 4% initial enrichment, 45 GWD/MTU burnup, and 5 year cooling time is {approx} 9.8 x 10{sup 4} counts per second (c/s) with the 100-Hz repeated interrogation pattern of 0 to 10 {micro}s interrogation, 0.2 ms to 1 ms counting time, and 1 x 10{sup 9} n/s neutron source. The {sup 244}Cm neutron background count rate for this counting time scheme is {approx} 1 x 10{sup 4} c/s, and thus the signal to background ratio is {approx}10.

  10. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    NASA Astrophysics Data System (ADS)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

    2015-07-01

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  11. Performance of an experimental polyethylene solid fuel ramjet

    NASA Astrophysics Data System (ADS)

    Gobbo Ferreira, J.; Silva, M. G.; de Carvalho, J. A.

    1999-09-01

    An experimental investigation on the characteristics of polyethylene combustion in a solid fuel ramjet is reported. An experimental expression for the fuel regression rate is presented for the conditions of the tests: r=2.78×10 -3p0.8cG0.35airT0.36air, where the units of ṙ are mm s -1, pc is the chamber pressure (atm), Gair is the air mass flux (g cm -2 s -1) and Tair is the air inlet temperature (K). Additionally, the effects of the introduction of an intermediate mixing chamber and of the increase in the post mixing chamber length over the combustion efficiency are studied. Radiation loss transfer rates are also analysed and discussed. The parameter was obtained by the use of a translucid polyethylene grain, which was first operated bare and later covered with an insulation aluminium sheet. Experimental results trends agree with those of previous works conducted with systems of similar geometry and different fuels.

  12. Experimental evaluation of a small fusion fuel cleanup systems

    SciTech Connect

    Holtslander, W.J.; Johnson, R.E.; Gravelle, F.B.; Schultz, C.M.

    1986-01-01

    Small tritium-burning experimental tokamaks will require some means of handling and purifying the deuterium-tritium fuel. A simple purification system would allow reinjection of fuel, minimize tritium inventory on site, and reduce the number of shipments of tritium to and from the tokamak site. This could simplify the licensing and safety aspects for sites unsuited to large inventories of tritium. At the request of the Canadian Fusion Fuels Technology Project, a number of conceptual designs of fusion fuel cleanup systems were prepared. The preferred design consisted of a gas circulation loop comprising an expansion tank, a pump, and a number of purification units, a uranium bed, a zirconium-aluminum getter bed, and two catalyst beds, Pt/Pd and CuO/MnO/sub 2/. This paper summarizes an experimental evaluation of this system using hydrogen and nontriated impurities. Using the information generated in the first part of the study, a simplified cleanup system containing two alternative purification paths was built and tested. The first path was through two uranium beds in series operating at 25 and 400/sup 0/C. In the second path, a zirconium-aluminum getter bed at 700/sup 0/C replaced the hot uranium bed. Both systems were demonstrated to be effective in the cleanup of a multicomponent gas mixture. These results show it is possible to have a simple cleanup system that is effective for purification of hydrogen that is typical of a fusion fuel mixture. This system provides for tritium recovery from the impurities, as well as purification.

  13. An experimental assembly for precise measurement of thermal accommodation coefficients

    NASA Astrophysics Data System (ADS)

    Trott, Wayne M.; Castañeda, Jaime N.; Torczynski, John R.; Gallis, Michael A.; Rader, Daniel J.

    2011-03-01

    An experimental apparatus has been developed to determine thermal accommodation coefficients for a variety of gas-surface combinations. Results are obtained primarily through measurement of the pressure dependence of the conductive heat flux between parallel plates separated by a gas-filled gap. Measured heat-flux data are used in a formula based on Direct Simulation Monte Carlo (DSMC) simulations to determine the coefficients. The assembly also features a complementary capability for measuring the variation in gas density between the plates using electron-beam fluorescence. Surface materials examined include 304 stainless steel, gold, aluminum, platinum, silicon, silicon nitride, and polysilicon. Effects of gas composition, surface roughness, and surface contamination have been investigated with this system; the behavior of gas mixtures has also been explored. Without special cleaning procedures, thermal accommodation coefficients for most materials and surface finishes were determined to be near 0.95, 0.85, and 0.45 for argon, nitrogen, and helium, respectively. Surface cleaning by in situ argon-plasma treatment reduced coefficient values by up to 0.10 for helium and by ˜0.05 for nitrogen and argon. Results for both single-species and gas-mixture experiments compare favorably to DSMC simulations.

  14. Computed isotopic inventory and dose assessment for SRS fuel and target assemblies

    SciTech Connect

    Chandler, M.C.; Ketusky, E.T.; Thoman, D.C.

    1995-06-19

    Past studies have identified and evaluated important radionuclide contributors to dose from reprocessed spent fuel sent to waste for Mark 16B and 22 fuel assemblies and for Mark 31 A and 31B target assemblies. Fission-product distributions after a 5- and 15-year decay time were calculated for a ``representative`` set of irradiation conditions (i.e., reactor power, irradiation time, and exposure) for each type of assembly. The numerical calculations were performed using the SHIELD/GLASS system of codes. The sludge and supernate source terms for dose were studied separately with the significant radionuclide contributors for each identified and evaluated. Dose analysis considered both inhalation and ingestion pathways: The inhalation pathway was analyzed for both evaporative and volatile releases. Analysis of evaporative releases utilized release fractions for the individual radionuclides as defined in the ICRP-30 by DOE guidance. A release fraction of unity was assumed for each radionuclide under volatile-type releases, which would encompass internally initiated events (e.g., fires, explosions), process-initiated events, and externally initiated events. Radionuclides which contributed at least 1% to the overall dose were designated as significant contributors. The present analysis extends and complements the past analyses through considering a broader spectrum of fuel types and a wider range of irradiation conditions. The results provide for a more thorough understanding of the influences of fuel composition and irradiation parameters on fission product distributions (at 2 years or more). Additionally, the present work allows for a more comprehensive evaluation of radionuclide contributions to dose and an estimation of the variability in the radionuclide composition of the dose source term that results from the spent fuel sent to waste encompassing a broad spectrum of fuel compositions and irradiation conditions.

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

    SciTech Connect

    Shore, Lawrence; Matlin, Ramail; Heinz, Robert

    2012-06-26

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

  16. Container for reprocessing and permanent storage of spent nuclear fuel assemblies

    DOEpatents

    Forsberg, Charles W.

    1992-01-01

    A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

  17. Experience gained from carrying out ultrasonic cleaning of fuel assemblies and control and protection system assemblies in the Novovoronezh NPP unit 3

    NASA Astrophysics Data System (ADS)

    Gorburov, V. I.; Shvarov, V. A.; Vitkovskii, S. L.

    2014-02-01

    A growth of deposits on fuel assembly elements was revealed during operation of the Novovoronezh NPP Unit 3 starting from 1997. This growth caused progressive reduction of coolant flow rate through the reactor core and increase of pressure difference across the assemblies, which eventually led to the need to reduce the power unit output and then to shut down the power unit. In view of these circumstances, it was decided to develop an installation for ultrasonic cleaning of fuel assemblies. The following conclusions were drawn with regard of this installation after completion of all stages of its development, commissioning, and improvement: no detrimental effect of ultrasound on the integrity of fuel assemblies was revealed, whereas the cleaning effect on the fuel assemblies subjected to ultrasonic treatment and improvement of their thermal-hydraulic characteristics are obvious. With these measures implemented, it became possible to clean all fuel assemblies in the core in 2011, to achieve better thermal-hydraulic characteristics, and to avoid reduction of power output and off-scheduled outages of Unit 3.

  18. Assessing the Feasibility of Using Neutron Resonance Transmission Analysis (NRTA) for Assaying Plutonium in Spent Fuel Assemblies

    SciTech Connect

    D. L. Chichester; J. W. Sterbentz

    2012-07-01

    Neutron resonance transmission analysis (NRTA) is an active-interrogation nondestructive assay (NDA) technique capable of assaying spent nuclear fuel to determine plutonium content. Prior experimental work has definitively shown the technique capable of assaying plutonium isotope composition in spent-fuel pins to a precision of approximately 3%, with a spatial resolution of a few millimeters. As a Grand Challenge to investigate NDA options for assaying spent fuel assemblies (SFAs) in the commercial fuel cycle, Idaho National Laboratory has explored the feasibility of using NRTA to assay plutonium in a whole SFA. The goal is to achieve a Pu assay precision of 1%. The NRTA technique uses low-energy neutrons from 0.1-40 eV, at the bottom end of the actinide-resonance range, in a time-of-flight arrangement. Isotopic composition is determined by relating absorption of the incident neutrons to the macroscopic cross-section of the actinides of interest in the material, and then using this information to determine the areal density of the isotopes in the SFA. The neutrons used for NRTA are produced using a pulsed, accelerator-based neutron source. Distinguishable resonances exist for both the plutonium (239,240,241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Based on extensive modeling of the problem using Monte Carlo-based simulation codes, our preliminary results suggest that by rotating an SFA to acquire four symmetric views, sufficient neutron transmission can be achieved to assay a SFA. In this approach multiple scan information for the same pins may also be unfolded to potentially allow the determination of plutonium for sub-regions of the assembly. For a 17 ? 17 pressurized water reactor SFA, a simplistic preliminary

  19. Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system

    DOEpatents

    Christiansen, David W.; Smith, Bob G.

    1982-01-01

    A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

  20. Influence of void fraction in the power distribution for a GE-12 fuel assembly

    NASA Astrophysics Data System (ADS)

    Castillo, S. J.; Vargas, G. A.; del Valle Gallegos, E.

    2017-01-01

    Analysis of the influence of void fraction distribution is very important to learn about the fission process and heat produced by the fuel assembly, here in this study several void fraction (VF) values along different burnup values have been considered in order to observe their influence in power distribution, uranium consumption, neutron flux andbehaviour for a GE-12 fuel assembly. For this study, burnups up to 60 MWd/kg and VF values up to 0.8 were considered setting the uranium enrichment at 3.5 weight percent at the start of every VF scenario, results show that higher void fractions reduce the thermal flux decreasing thermal fission and limiting heat production.

  1. A comparison of spent fuel assembly control instruments: The Cadarache PYTHON and the Los Alamos Fork

    SciTech Connect

    Bignan, G.; Capsie, J.; Romeyer-Dherbey, J. . Direction des Reacteurs Nucleaires); Rinard, P. )

    1991-01-01

    Devices to monitor spent fuel assemblies while stored under water with nondestructive assay methods, have been developed in France and in the United States. Both devices are designed to verify operator's declared values of exposures and cooling-time but the applications and thus the designs of the systems differ. A study, whose results are presented in this paper, has been conducted to compare the features and the performances of the two instruments. 4 refs., 9 figs.

  2. Economic incentives for additional critical experimentation applicable to fuel dissolution

    SciTech Connect

    Mincey, J.F.; Primm, R.T. III; Waltz, W.R.

    1981-01-01

    Fuel dissolution operations involving soluble absorbers for criticality control are among the most difficult to establish economical subcritical limits. The paucity of applicable experimental data can significantly hinder a precise determination of a bias in the method chosen for calculation of the required soluble absorber concentration. Resorting to overly conservative bias estimates can result in excessive concentrations of soluble absorbers. Such conservatism can be costly, especially if soluble absorbers are used in a throw-away fashion. An economic scoping study is presented which demonstrates that additional critical experimentation will likely lead to reductions in the soluble absorber (i.e., gadolinium) purchase costs for dissolution operations. The results indicate that anticipated savings maybe more than enough to pay for the experimental costs.

  3. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-08-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  4. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-01-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  5. Iterative ct reconstruction from few projections for the nondestructive post irradiation examination of nuclear fuel assemblies

    NASA Astrophysics Data System (ADS)

    Abir, Muhammad Imran Khan

    The core components (e.g. fuel assemblies, spacer grids, control rods) of the nuclear reactors encounter harsh environment due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of the nuclear power plants. The Post Irradiation Examination (PIE) can reveal information about the integrity of the elements during normal operations and off?normal events. Computed tomography (CT) is a tool for evaluating the structural integrity of elements non-destructively. CT requires many projections to be acquired from different view angles after which a mathematical algorithm is adopted for reconstruction. Obtaining many projections is laborious and expensive in nuclear industries. Reconstructions from a small number of projections are explored to achieve faster and cost-efficient PIE. Classical reconstruction algorithms (e.g. filtered back projection) cannot offer stable reconstructions from few projections and create severe streaking artifacts. In this thesis, conventional algorithms are reviewed, and new algorithms are developed for reconstructions of the nuclear fuel assemblies using few projections. CT reconstruction from few projections falls into two categories: the sparse-view CT and the limited-angle CT or tomosynthesis. Iterative reconstruction algorithms are developed for both cases in the field of compressed sensing (CS). The performance of the algorithms is assessed using simulated projections and validated through real projections. The thesis also describes the systematic strategy towards establishing the conditions of reconstructions and finds the optimal imaging parameters for reconstructions of the fuel assemblies from few projections.

  6. An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

    SciTech Connect

    Swinhoe, Martyn T; Tobin, Stephen J; Fensin, Mike L; Menlove, Howard O

    2009-01-01

    be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC&A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument will be quantified for the full assembly

  7. Fuel cell assembly unit for promoting fluid service and electrical conductivity

    DOEpatents

    Jones, Daniel O.

    1999-01-01

    Fluid service and/or electrical conductivity for a fuel cell assembly is promoted. Open-faced flow channel(s) are formed in a flow field plate face, and extend in the flow field plate face between entry and exit fluid manifolds. A resilient gas diffusion layer is located between the flow field plate face and a membrane electrode assembly, fluidly serviced with the open-faced flow channel(s). The resilient gas diffusion layer is restrained against entering the open-faced flow channel(s) under a compressive force applied to the fuel cell assembly. In particular, a first side of a support member abuts the flow field plate face, and a second side of the support member abuts the resilient gas diffusion layer. The support member is formed with a plurality of openings extending between the first and second sides of the support member. In addition, a clamping pressure is maintained for an interface between the resilient gas diffusion layer and a portion of the membrane electrode assembly. Preferably, the support member is spikeless and/or substantially flat. Further, the support member is formed with an electrical path for conducting current between the resilient gas diffusion layer and position(s) on the flow field plate face.

  8. Implementation of the active neutron Coincidence Collar for the verification of unirradiated PWR and BWR fuel assemblies

    SciTech Connect

    Menlove, H.O.; Keddar, A.

    1982-01-01

    An active neutron interrogation technique has been developed for the measurement of the /sup 235/U content in fresh fuel assemblies. The method employs an AmLi neutron source to induce fission reactions in the fuel assembly and coincidence counting of the resulting fission reaction neutrons. When no interrogation source is present, the passive neutron coincidence rate gives a measure of the /sup 238/U by the spontaneous fission reactions. The system can be applied to the fissile content determination in fresh fuel assemblies for accountability, criticality control, and safeguards purposes. Field tests have been performed by International Atomic Energy Agency (IAEA) staff using the Coincidence Collar to verify the /sup 235/U content in light-water-reactor fuel assemblies. The results gave an accuracy of 1 to 2% in the active mode (/sup 235/U) and 2 to 3% in the passive mode (/sup 238/U) under field conditions.

  9. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    SciTech Connect

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  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. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  12. Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond

    DOEpatents

    Phillips, John R.; Halbig, James K.; Menlove, Howard O.; Klosterbuer, Shirley F.

    1985-01-01

    A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

  13. Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond

    DOEpatents

    Phillips, J.R.; Halbig, J.K.; Menlove, H.O.; Klosterbuer, S.F.

    1984-01-01

    A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

  14. Experimental results with hydrogen fueled internal combustion engines

    NASA Technical Reports Server (NTRS)

    De Boer, P. C. T.; Mclean, W. J.; Homan, H. S.

    1975-01-01

    The paper focuses on the most important experimental findings for hydrogen-fueled internal combustion engines, with particular reference to the application of these findings to the assessment of the potential of hydrogen engines. Emphasis is on the various tradeoffs that can be made, such as between maximum efficiency, maximum power, and minimum NO emissions. The various possibilities for induction and ignition are described. Some projections are made about areas in which hydrogen engines may find their initial application and about optimum ways to design such engines. It is shown that hydrogen-fueled reciprocal internal combustion engines offer important advantages with respect to thermal efficiency and exhaust emissions. Problems arising from preignition can suitably be avoided by restricting the fuel-air equivalence ratio to values below about 0.5. The direct cylinder injection appears to be a very attractive way to operate the engine, because it combines a wide range of possible power outputs with a high thermal efficiency and very low NO emissions at part loads.

  15. Prevention of significant deterioration permit application for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    SciTech Connect

    Not Available

    1989-08-01

    This New Source Review'' has been submitted by the US Department of Energy-Richland Operations Office (PO Box 550, Richland, Washington 99352), pursuant to WAC 173-403-050 and in compliance with the Department of Ecology Guide to Processing A Prevention Of Significant Deterioration (PSD) Permit'' for three new sources of radionuclide emissions at the Hanford Site in Washington State. The three new sources, the Fueled Clad Fabrication System (FCFS), the Radioisotope Power Systems Facility (RPSF), and the Fuel Assembly Area (FAA), will be located in one facility, the Fuels and Materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post-irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies for use in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, these systems will be dealt with separately to the extent possible. The FAA is a comparatively independent operation though it will share the FMEF complex.

  16. Supplemental information for a notice of construction for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    SciTech Connect

    Not Available

    1989-08-01

    This ''Notice of Construction'' has been submitted by the US Department of Energy-Richland Operations Office (P.O. Box 550, Richland, Washington 99352), pursuant to WAC 402-80-070, for three new sources of radionuclide emissions at the Hanford Site in Washington State (Figure 1). The three new sources, the Fueled Clad Fabrication System (FCFS) the Radioisotope Power Systems Facility (RPSF) and the Fuel Assembly Area (FAA) will be located in one facility, the Fuels and materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post- irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies to be used in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, to the extent possible, these systems will be dealt with separately. The FAA is a comparatively independent operation though it will share the FMEF complex.

  17. Application for approval for construction of the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    SciTech Connect

    Not Available

    1989-08-01

    The following ''Application for Approval of Construction'' is being submitted by the US Department of Energy-Richland Operations Office, pursuant to 40 CFR 61.07, for three new sources of airborne radionuclide emissions at the Hanford Site in Washington State. The three new sources, the Fueled Clad Fabrication System (FCFS), the Radioisotope Power Systems Facility (RPSF), and the Fuel Assembly Area (FAA), will be located in one facility, the Fuels and Materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post-irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were canceled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies to be used in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building and stack and, in certain cases, the same floor space. Given this relationship, these systems will be dealt with separately to the extent possible. The FAA is a comparatively independent operation though it will share the FMEF complex. 2 refs., 16 figs., 12 tabs.

  18. Controlling the hydrophilicity and contact resistance of fuel cell bipolar plate surfaces using layered nanoparticle assembly

    NASA Astrophysics Data System (ADS)

    Wang, Feng

    Hybrid nanostructured coatings exhibiting the combined properties of electrical conductivity and surface hydrophilicity were obtained by using Layer-by-Layer (LBL) assembly of cationic polymer, silica nanospheres, and carbon nanoplatelets. This work demonstrates that by controlling the nanoparticle zeta (zeta) potential through the suspension parameters (pH, organic solvent type and amount, and ionic content) as well as the assembly sequence, the nanostructure and composition of the coatings may be adjusted to optimize the desired properties. Two types of silica nanospheres were evaluated as the hydrophilic component: X-TecRTM 3408 from Nano-X Corporation, with a diameter of about 20 nm, and polishing silica from Electron Microscopy Supply, with diameter of about 65 nm. Graphite nanoplatelets with a thickness of 5~10nm (Aquadag RTM E from Acheson Industries) were used as electrically conductive filler. A cationic copolymer of acrylamide and a quaternary ammonium salt (SuperflocRTM C442 from Cytec Corporation) was used as the binder for the negatively charged nanoparticles. Coatings were applied to gold-coated stainless steel substrates presently used a bipolar plate material for proton exchange membrane (PEM) fuel cells. Coating thickness was found to vary nearly linearly with the number of polymer-nanoparticle layers deposited while a monotonic increase in coating contact resistance was observed for all heterogeneous and pure silica coatings. Thickness increased if the difference in the oppositely charged zeta potentials of the adsorbing components was enhanced through alcohol addition. Interestingly, an opposite effect was observed if the zeta potential difference was increased through pH variation. This previously undocumented difference in adsorption behavior is herein related to changes to the surface chemical heterogeneity of the nanoparticles. Coating contact resistance and surface wettability were found to have a more subtle dependence on the assembly

  19. Mechanical and thermomechanical calculations related to the storage of spent nuclear-fuel assemblies in granite

    SciTech Connect

    Butkovich, T.R.

    1981-08-01

    A generic test of the geologic storage of spent-fuel assemblies from an operating nuclear reactor is being made by the Lawrence Livermore National Laboratory at the US Department of Energy`s Nevada Test Site. The spent-fuel assemblies were emplaced at a depth of 420 m (1370 ft) below the surface in a typical granite and will be retrieved at a later time. The early time, close-in thermal history of this type of repository is being simulated with spent-fuel and electrically heated canisters in a central drift, with auxiliary heaters in two parallel side drifts. Prior to emplacement of the spent-fuel canisters, preliminary calculations were made using a pair of existing finite-element codes. Calculational modeling of a spent-fuel repository requires a code with a multiple capability. The effects of both the mining operation and the thermal load on the existing stress fields and the resultant displacements of the rock around the repository must be calculated. The thermal loading for each point in the rock is affected by heat tranfer through conduction, radiation, and normal convection, as well as by ventilation of the drifts. Both the ADINA stress code and the compatible ADINAT heat-flow code were used to perform the calculations because they satisfied the requirements of this project. ADINAT was adapted to calculate radiative and convective heat transfer across the drifts and to model the effects of ventilation in the drifts, while the existing isotropic elastic model was used with the ADINA code. The results of the calculation are intended to provide a base with which to compare temperature, stress, and displacement data taken during the planned 5-y duration of the test. In this way, it will be possible to determine how the existing jointing in the rock influences the results as compared with a homogeneous, isotropic rock mass. Later, new models will be introduced into ADINA to account for the effects of jointing.

  20. Hanford MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site (SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. Hanford has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 1 facility. In all, a total of three LA MOX fuel fabrication options were identified by Hanford that could accommodate the program. In every case, only minor modification would be required to ready any of the facilities to accept the equipment necessary to accomplish the LA program.

  1. Expert System analysis of non-fuel assembly hardware and spent fuel disassembly hardware: Its generation and recommended disposal

    SciTech Connect

    Williamson, D.A.

    1991-12-31

    Almost all of the effort being expended on radioactive waste disposal in the United States is being focused on the disposal of spent Nuclear Fuel, with little consideration for other areas that will have to be disposed of in the same facilities. one area of radioactive waste that has not been addressed adequately because it is considered a secondary part of the waste issue is the disposal of the various Non-Fuel Bearing Components of the reactor core. These hardware components fall somewhat arbitrarily into two categories: Non-Fuel Assembly (NFA) hardware and Spent Fuel Disassembly (SFD) hardware. This work provides a detailed examination of the generation and disposal of NFA hardware and SFD hardware by the nuclear utilities of the United States as it relates to the Civilian Radioactive Waste Management Program. All available sources of data on NFA and SFD hardware are analyzed with particular emphasis given to the Characteristics Data Base developed by Oak Ridge National Laboratory and the characterization work performed by Pacific Northwest Laboratories and Rochester Gas & Electric. An Expert System developed as a portion of this work is used to assist in the prediction of quantities of NFA hardware and SFD hardware that will be generated by the United States` utilities. Finally, the hardware waste management practices of the United Kingdom, France, Germany, Sweden, and Japan are studied for possible application to the disposal of domestic hardware wastes. As a result of this work, a general classification scheme for NFA and SFD hardware was developed. Only NFA and SFD hardware constructed of zircaloy and experiencing a burnup of less than 70,000 MWD/MTIHM and PWR control rods constructed of stainless steel are considered Low-Level Waste. All other hardware is classified as Greater-ThanClass-C waste.

  2. Single-wall carbon nanotube-based proton exchange membrane assembly for hydrogen fuel cells.

    PubMed

    Girishkumar, G; Rettker, Matthew; Underhile, Robert; Binz, David; Vinodgopal, K; McGinn, Paul; Kamat, Prashant

    2005-08-30

    A membrane electrode assembly (MEA) for hydrogen fuel cells has been fabricated using single-walled carbon nanotubes (SWCNTs) support and platinum catalyst. Films of SWCNTs and commercial platinum (Pt) black were sequentially cast on a carbon fiber electrode (CFE) using a simple electrophoretic deposition procedure. Scanning electron microscopy and Raman spectroscopy showed that the nanotubes and the platinum retained their nanostructure morphology on the carbon fiber surface. Electrochemical impedance spectroscopy (EIS) revealed that the carbon nanotube-based electrodes exhibited an order of magnitude lower charge-transfer reaction resistance (R(ct)) for the hydrogen evolution reaction (HER) than did the commercial carbon black (CB)-based electrodes. The proton exchange membrane (PEM) assembly fabricated using the CFE/SWCNT/Pt electrodes was evaluated using a fuel cell testing unit operating with H(2) and O(2) as input fuels at 25 and 60 degrees C. The maximum power density obtained using CFE/SWCNT/Pt electrodes as both the anode and the cathode was approximately 20% better than that using the CFE/CB/Pt electrodes.

  3. Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

    SciTech Connect

    Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh; Ba Vien Luong; Kien Cuong Nguyen

    2008-07-15

    The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configuration with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)

  4. Experimental plan for the fuel-oil study

    SciTech Connect

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

    1992-01-01

    An up-to-date assessment of the Weatherization Assistance Program (WAP) is being performed by the US Department of Energy WAP Division and the Oak Ridge National Laboratory. Five studies form the evaluation. Major goals of the Fuel-Oil Study are to estimate the fuel oil saved by the WAP in the Northeast during the 1990 and 1991 program years, identify and quantify non-energy impacts of the WAP, assess the cost effectiveness of the WAP within this submarket, and assess factors which may cause savings and cost effectiveness to vary. The study will only analyze single-family houses in the nine states in the Northeast census region and will be carried out over two heating seasons (1990 and 1991 WAP program years). A split-winter, pre- and post-weatherization experimental design with a control group will be used. Houses will be monitored over one winter. Energy conservation measures will be installed in the weatherized houses in January of each winter by the local WAP subgrantee. One hundred twenty five weatherized houses and 75 control houses will be monitored over the 1990--1991 winter; a different set of 200 houses will be monitored over the 1991--1992 winter. The houses will be evenly distributed among 25 subgrantees. Space-heating fuel-oil consumption, indoor temperature, and outdoor temperature data will be collected for all houses. Fuel-oil delivery data will be collected for each house monitored over the 1990--1991 winter for at least a year before weatherization. The delivery data will be analyzed to determine if the accuracy of the study can be improved by collecting fuel-oil delivery data on a larger sample of houses over the 1991--1992 winter. Detailed survey information will be obtained on all the houses. This information includes descriptive details of the house and its mechanical systems, details on household size and other demographics, and occupant answers to questions regarding comfort, safety, and operation of their space-heating system and house.

  5. SRS MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO{sub 2} and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program

  6. Development of coupled SCALE4.2/GTRAN2 computational capability for advanced MOX fueled assembly designs

    SciTech Connect

    Vujic, J.; Greenspan, E.; Slater, Postma, T.; Casher, G.; Soares, I.; Leal, L.

    1995-05-01

    An advanced assembly code system that can efficiently and accurately analyze various designs (current and advanced) proposed for plutonium disposition is being developed by {open_quotes}marrying{close_quotes} two existing state-of-the-art methodologies-GTRAN2 and SCALE 4.2. The resulting code system, GT-SCALE, posses several unique characteristics: exact 2D representation of a complete fuel assembly, while preserving the heterogeniety of each of its pin cells; flexibility in the energy group structure, the present upper limit being 218 groups; a comprehensive cross-section library and material data base; and accurate burnup calculations. The resulting GT-SCALE is expected to be very useful for a wide variety of applications, including the analysis of very heterogeneous UO{sub 2} fueled LWR fuel assemblies; of hexagonal shaped fuel assemblies as of the Russian LWRs; of fuel assemblies for HTGRs; as well as for the analysis of criticality safety and for calculation of the source term of spent fuel.

  7. Neutronic performance of several LEU fuel assembly designs for the WWR-SM research reactor in Uzbekistan.

    SciTech Connect

    Bretscher, M. M.; Hanan, N. A.; Matos, J. E.; Yuldashev, B. S.; Baytelesov, S.; Rakhmanov, A.; Technology Development; Inst. of Nuclear Physics

    2002-01-01

    The 10 MW WWR-SM research reactor in Uzbekistan currently uses HEU (36%) IRT-3M 6-tube fuel assemblies manufactured by the Novosibirsk Chemical Concentrates Plant in Russia. Recent 4x4 core configurations reflected by beryllium have been operated at 8 MW. The Institute of Nuclear Physics plans to convert the reactor to LEU (19.7%) fuel as soon as a suitable LEU fuel assembly is qualified. This study compares the neutronic performance of the reactor and its experiments using LEU pin-type and LEU tube-type fuel assembly designs with the current HEU (36%) reference fuel assembly. Both 3D Monte Carlo and 3D diffusion theory calculations were performed to analyze a critical core configuration with partially-burned HEU fuel assemblies in order to establish the credibility of the analytical methods and computer models used to describe the reactor and its experiments. Results based on these techniques are in reasonable agreement with the measured data. An LEU pin-type design (164 pins, 4.5 g U/cm{sup 3}, 375g {sup 235}U) or an LEU tube-type design (IRT-3M, 6-tube, 5.4 gU/cm{sup 3}, and 364g {sup 235}U) with U9Mo-Al fuel meat could operate with about the same cycle length and experiment load as the reference HEU (36%) IRT-3M fuel. The annual fuel assembly consumption would be nearly the same in these HEU and LEU cores. For the LEU pin-type design, fast (thermal) fluxes would be reduced by 2.5% (14%) for experiments located at the center of the fuel assemblies and by 0.5% (4%) for experiments located in experiment channels in the beryllium reflector. For the LEU tube-type design, fast (thermal) fluxes would be reduced by 3.5% (15%) for experiments located at the center of the fuel assemblies and by 1.2% (5%) for experiments located in experiment channels in the beryllium reflector. If the {sup 235}U content of the LEU pin-type fuel assemblies were increased to 480g (using pins similar to those planned to be tested in the WWR-M reactor at Gatchina, Russia in 2003 and 2004

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

    SciTech Connect

    Nix, Andrew Carl

    2015-03-23

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

  9. Optimizing membrane electrode assembly of direct methanol fuel cells for portable power

    NASA Astrophysics Data System (ADS)

    Liu, Fuqiang

    Direct methanol fuel cells (DMFCs) for portable power applications require high power density, high-energy conversion efficiency and compactness. These requirements translate to fundamental properties of high methanol oxidation and oxygen reduction kinetics, as well as low methanol and water crossover. In this thesis a novel membrane electrode assembly (MEA) for direct methanol fuel cells has been developed, aiming to improve these fundamental properties. Firstly, methanol oxidation kinetics has been enhanced and methanol crossover has been minimized by proper control of ionomer crystallinity and its swelling in the anode catalyst layer through heat-treatment. Heat-treatment has a major impact on anode characteristics. The short-cured anode has low ionomer crystallinity, and thus swells easily when in contact with methanol solution to create a much denser anode structure, giving rise to higher methanol transport resistance than the long-cured anode. Variations in interfacial properties in the anode catalyst layer (CL) during cell conditioning were also characterized, and enhanced kinetics of methanol oxidation and severe limiting current phenomenon were found to be caused by a combination of interfacial property variations and swelling of ionomer over time. Secondly, much effort has been expended to develop a cathode CL suitable for operation under low air stoichiometry. The effects of fabrication procedure, ionomer content, and porosity distribution on the microstructure and cathode performance under low air stoichiometry are investigated using electrochemical and surface morphology characterizations to reveal the correlation between microstructure and electrochemical behavior. At the same time, computational fluid dynamics (CFD) models of DMFC cathodes have been developed to theoretically interpret the experimental results, to investigate two-phase transport, and to elucidate mechanism of cathode mixed potential due to methanol crossover. Thirdly, a MEA with low

  10. Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

    SciTech Connect

    Durkee, Jr., Joe W.

    2015-06-19

    A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20, 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/137Cs 134Cs/154Eu, and 154Eu/137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the

  11. A general mathematical model for analyzing the performance of fuel-cell membrane-electrode assemblies

    NASA Astrophysics Data System (ADS)

    Zhu, Huayang; Kee, Robert J.

    We have developed a general mathematical model to represent the membrane-electrode assembly (MEA) of fuel-cell systems. The model is used to analyze the effects of various polarization resistances on cell performance. The model accommodates arbitrary gas mixtures on the anode and cathode sides of the MEA. Moreover, it accommodates a variety of porous electrode and electrolyte structures. Concentration overpotentials are based on a dusty-gas representation of transport through porous electrodes. The activation overpotentials are represented using the Butler-Volmer equation. Although the model is general, the emphasis in this paper is on solid-oxide fuel-cell (SOFC) systems for the direct electrochemical oxidation (DECO) of hydrocarbons.

  12. Hydraulically actuated fuel injector including a pilot operated spool valve assembly and hydraulic system using same

    DOEpatents

    Shafer, Scott F.

    2002-01-01

    The present invention relates to hydraulic systems including hydraulically actuated fuel injectors that have a pilot operated spool valve assembly. One class of hydraulically actuated fuel injectors includes a solenoid driven pilot valve that controls the initiation of the injection event. However, during cold start conditions, hydraulic fluid, typically engine lubricating oil, is particularly viscous and is often difficult to displace through the relatively small drain path that is defined past the pilot valve member. Because the spool valve typically responds slower than expected during cold start due to the difficulty in displacing the relatively viscous oil, accurate start of injection timing can be difficult to achieve. There also exists a greater difficulty in reaching the higher end of the cold operating speed range. Therefore, the present invention utilizes a fluid evacuation valve to aid in displacement of the relatively viscous oil during cold start conditions.

  13. Container for reprocessing and permanent storage of spent nuclear fuel assemblies

    DOEpatents

    Forsberg, C.W.

    1992-03-24

    A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

  14. Main experimental results of the Phebus Severe Fuel Damage Program

    SciTech Connect

    Gonnier, C. )

    1993-01-01

    The main objective of this program is to improve our knowledge about the early phase of a pressurized water reactor in-vessel core melt degradation in a temperature range up to 2800 K. The experimental program performed from December 1986 to June 1989 consists of six in-pile experiments with 21 fresh fuel rod bundles of 0.8 m active length. It is divided into two series of tests: (1) The first one (B9, B9R, B9+) is mainly devoted to the oxidation phenomenon and its consequences for fuel degradation. This series is characterized by high oxidation rates. (2) The second series [C3, C3+, Ag-In-CD (AIC)] is characterized by low oxidation rates of the cladding in order to study the interaction between the remaining Zircaloy and the other materials: interactions with Inconel and UO[sub 2] for C3 and C3+ tests and interactions with the Ag-In-Cd alloy and stainless steel of the control rod for the AIC test.

  15. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    SciTech Connect

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G.; Carrell, R.D.; Jaeger, C.D.; Thompson, M.L.; Strasser, A.A.

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET.

  16. Performance and microbial ecology of air-cathode microbial fuel cells with layered electrode assemblies.

    PubMed

    Butler, Caitlyn S; Nerenberg, Robert

    2010-05-01

    Microbial fuel cells (MFCs) can be built with layered electrode assemblies, where the anode, proton exchange membrane (PEM), and cathode are pressed into a single unit. We studied the performance and microbial community structure of MFCs with layered assemblies, addressing the effect of materials and oxygen crossover on the community structure. Four MFCs with layered assemblies were constructed using Nafion or Ultrex PEMs and a plain carbon cloth electrode or a cathode with an oxygen-resistant polytetrafluoroethylene diffusion layer. The MFC with Nafion PEM and cathode diffusion layer achieved the highest power density, 381 mW/m(2) (20 W/m(3)). The rates of oxygen diffusion from cathode to anode were three times higher in the MFCs with plain cathodes compared to those with diffusion-layer cathodes. Microsensor studies revealed little accumulation of oxygen within the anode cloth. However, the abundance of bacteria known to use oxygen as an electron acceptor, but not known to have exoelectrogenic activity, was greater in MFCs with plain cathodes. The MFCs with diffusion-layer cathodes had high abundance of exoelectrogenic bacteria within the genus Geobacter. This work suggests that cathode materials can significantly influence oxygen crossover and the relative abundance of exoelectrogenic bacteria on the anode, while PEM materials have little influence on anode community structure. Our results show that oxygen crossover can significantly decrease the performance of air-cathode MFCs with layered assemblies, and therefore limiting crossover may be of particular importance for these types of MFCs.

  17. Fuel cleanup system for the tritium systems test assembly: design and experiments

    SciTech Connect

    Kerr, E.C.; Bartlit, J.R.; Sherman, R.H.

    1980-01-01

    A major subsystem of the Tritium Systems Test Assembly is the Fuel Cleanup System (FCU) whose functons are to: (1) remove impurities in the form of argon and tritiated methane, water, and ammonia from the reactor exhaust stream and (2) recover tritium for reuse from the tritiated impurities. To do this, a hybrid cleanup system has been designed which utilizes and will test concurrently two differing technologies - one based on disposable, hot metal (U and Ti) getter beds and a second based on regenerable cryogenic asdorption beds followed by catalytic oxidation of impurities to DTO and stackable gases and freezout of the resultant DTO to recover essentially all tritium for reuse.

  18. Modeling and experimental diagnostics in polymer electrolyte fuel cells

    NASA Astrophysics Data System (ADS)

    Springer, T. E.; Wilson, M. S.; Gottesfeld, S.

    1993-12-01

    This paper presents a fit between model and experiments for well-humidified polymer electrolyte fuel cells operated to maximum current density with a range of cathode gas compositions. The model considers, in detail, losses caused by: (1) interfacial kinetics at the Pt/ionomer interface; (2) gas-transport and ionic-conductivity limitations in the catalyst layer; and (3) gas-transport limitations in the cathode backing. Our experimental data were collected with cells that utilized thin-film catalyst layers bonded directly to the membrane, and a separate catalyst-free hydrophobic backing layer. This structure allows a clearer resolution of the processes taking place in each of these distinguishable parts of the cathode. In our final comparison of model predictions with the experimental data, we stress the simultaneous fit of a family of complete polarization curves obtained for gas compositions ranging from 5 atoms O2 to a mixture of 5% O2 in N2, employing in each case the same model parameters for interracial kinetics, catalyst-layer transport, and backing-layer transport. This approach allowed us to evaluate losses in the cathode backing and in the cathode catalyst layer, and thus identify the improvements required to enhance the performance of air cathodes in polymer electrolyte fuel cells. Finally, we show that effects of graded depletion in oxygen along the gas flow channel can be accurately modeled using a uniform effective oxygen concentration in the flow channel, equal to the average of inlet and exit concentrations. This approach has enabled simplified and accurate consideration of oxygen utilization effects.

  19. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Maldonado, G I; Burdo, J; He, T

    2006-10-10

    A technical safeguards challenge has remained for decades for the IAEA to identify possible diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies. In fact, as modern nuclear power plants are pushed to higher power levels and longer fuel cycles, fuel failures (i.e., ''leakers'') as well as the corresponding fuel assembly repairs (i.e., ''reconstitutions'') are commonplace occurrences within the industry. Fuel vendors have performed hundreds of reconstitutions in the past two decades, thus, an evolved know-how and sophisticated tools exist to disassemble irradiated fuel assemblies and replace damaged pins with dummy stainless steel or other type rods. Various attempts have been made in the past two decades to develop a technology to identify a possible diversion of pin(s) and to determine whether some pins are missing or replaced with dummy or fresh fuel pins. However, to date, there are no safeguards instruments that can detect a possible pin diversion scenario to the requirements of the IAEA. The FORK detector system [1-2] can characterize spent fuel assemblies using operator declared data, but it is not sensitive enough to detect missing pins from spent fuel assemblies. Likewise, an emission computed tomography system [3] has been used to try to detect missing pins from a spent fuel assembly, which has shown some potential for identifying possible missing pins but this capability has not yet been fully demonstrated. The use of such a device in the future would not be envisaged, especially in an inexpensive, easy to handle setting for field applications. In this article, we describe a concept and ongoing research to help develop a new safeguards instrument for the detection of pin diversions in a PWR spent fuel assembly. The proposed instrument is based on one or more very thin radiation detectors that could be inserted within the guide tubes of a Pressurized Water Reactor (PWR) assembly. Ultimately, this work could lead to the

  20. Rail Shock and Vibration Pre-Test Modeling of a Used Nuclear Fuel Assembly

    SciTech Connect

    Ross, Steven B.; Klymyshyn, Nicholas A.; Jensen, Philip J.; Best, Ralph E.; Maheras, Steven J.; McConnell, Paul E.; Orchard, John

    2015-04-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste (HLW). The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and HLW generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC is responsible for addressing issues regarding the long-term or extended storage (ES) of UNF and its subsequent transportation. Available information is not sufficient to determine the ability of ES UNF, including high-burnup fuel, to withstand shock and vibration forces that could occur when the UNF is shipped by rail from nuclear power plant sites to a storage or disposal facility. There are three major gaps in the available information – 1) the forces that UNF assemblies would be subjected to when transported by rail, 2) the mechanical characteristics of fuel rod cladding, which is an essential structure for controlling the geometry of the UNF, a safety related feature, and 3) modeling methodologies to evaluate multiple possible degradation or damage mechanisms over the UNF lifetime. In order to address the first gap, options for tests to determine the physical response of surrogate UNF assemblies subjected to shock and vibration forces that are expected to be experienced during normal conditions of transportation (NCT) by rail must be identified and evaluated. The objective of the rail shock and vibration tests is to obtain data that will help researchers understand the mechanical loads that ES UNF assemblies would be subjected to under normal conditions of transportation and to fortify the computer modeling that will be necessary to evaluate the impact

  1. Void fraction distribution in a boiling water reactor fuel assembly and the evaluation of subchannel analysis codes

    SciTech Connect

    Inoue, Akira; Futakuchi, Masanobu; Yagi, Makoto; Mitsutake, Toru; Morooka, Shinichi

    1995-12-01

    Void fraction measurement tests for boiling water reactor (BWR) simulated nuclear fuel assemblies have been conducted using an X-ray computed tomography scanner.there are two types of fuel assemblies concerning water rods. One fuel assembly has two water rods; the other has one large water rod. The effects of the water rods on radial void fraction distributions are measured within the fuel assemblies. The results show that the water rod effect does not make a large difference in void fraction distribution. The subchannel analysis codes COBRA/BWR and THERMIT-2 were compared with subchannel-averaged void fractions. The prediction accuracy of COBRA/BWR and THERMIT-2 for the subchannel-averaged void fraction was {Delta}{alpha} = {minus}3.6%, {sigma} = 4.8% and {Delta}{alpha} = {minus}4.1%, {sigma} = 4.5%, respectively, where {Delta}{alpha} is the average of the difference measured and calculated values. The subchannel analysis codes are highly applicable for the prediction of a two-phase flow distribution within BWR fuel assemblies.

  2. Synthetic Molecular Machines for Active Self-Assembly: Prototype Algorithms, Designs, and Experimental Study

    NASA Astrophysics Data System (ADS)

    Dabby, Nadine L.

    Computer science and electrical engineering have been the great success story of the twentieth century. The neat modularity and mapping of a language onto circuits has led to robots on Mars, desktop computers and smartphones. But these devices are not yet able to do some of the things that life takes for granted: repair a scratch, reproduce, regenerate, or grow exponentially fast--all while remaining functional. This thesis explores and develops algorithms, molecular implementations, and theoretical proofs in the context of "active self-assembly" of molecular systems. The long-term vision of active self-assembly is the theoretical and physical implementation of materials that are composed of reconfigurable units with the programmability and adaptability of biology's numerous molecular machines. En route to this goal, we must first find a way to overcome the memory limitations of molecular systems, and to discover the limits of complexity that can be achieved with individual molecules. One of the main thrusts in molecular programming is to use computer science as a tool for figuring out what can be achieved. While molecular systems that are Turing-complete have been demonstrated [Winfree, 1996], these systems still cannot achieve some of the feats biology has achieved. One might think that because a system is Turing-complete, capable of computing "anything," that it can do any arbitrary task. But while it can simulate any digital computational problem, there are many behaviors that are not "computations" in a classical sense, and cannot be directly implemented. Examples include exponential growth and molecular motion relative to a surface. Passive self-assembly systems cannot implement these behaviors because (a) molecular motion relative to a surface requires a source of fuel that is external to the system, and (b) passive systems are too slow to assemble exponentially-fast-growing structures. We call these behaviors "energetically incomplete" programmable

  3. Measurements on spent-fuel assemblies at Arkansas Nuclear One using the Fork system. Final report, January 1995

    SciTech Connect

    Ewing, R.I.; Bronowski, D.R.; Bosler, G.E.; Siebelist, R.; Priore, J.; Hansford, C.H.; Sullivan, S.

    1997-03-01

    The Fork measurement system has been used to examine spent-fuel assemblies at the two reactors of Arkansas Nuclear One, operated by Entergy Operations, Inc. The Unit 1 reactor is a Babcock and Wilcox (B and W) design, and the Unit 2 reactor is a Combustion Engineering (CE) design. The neutron and gamma-ray emissions from individual spent-fuel assemblies were measured in the storage pools by raising each assembly pathway out of the storage rack and performing a measurement near the center of the assembly. The overall accuracy of the measurements after corrections is about 2%. Thirty-four assemblies were examined at Unit 1, and forty-one assemblies at Unit 2. The average deviation of the burnup measurements from the calibration was 3.0% at Unit 1 and 3.5% at Unit 2, indicating 2 to 3% random variation among the reactor records. There was no indication of clearly anomalous assemblies. Axial Scans of the variation in neutron and gamma ray emission were obtained by collecting data at several locations along the length of three assemblies at Unit 2. Two of these assemblies were nonstandard in that each contained a small neutron source. The sources were detected by the axial scans. The test program was a cooperative effort involving Sandia National Laboratories, Los Alamos National Laboratory, Entergy Operations, Inc., the Electric Power Research Institute, and the Office of Civilian Radioactive Waste Management of the US Department of Energy.

  4. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    SciTech Connect

    Bronson, M.C.

    1997-10-01

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium.

  5. Westinghouse Fuel Assemblies Performance after Operation in South-Ukraine NPP Mixed Core

    SciTech Connect

    Abdullayev, A. M.; Kulish, G. V.; Slyeptsov, O.; Slyeptsov, S.; Aleshin, Y.; Sparrow, S.; Lashevych, P.; Sokolov, D.; Latorre, Richard

    2013-09-14

    The evaluation of WWER-1000 Westinghouse fuel performance was done using the results of post–irradiation examinations of six LTAs and the WFA reload batches that have operated normally in mixed cores at South-Ukraine NPP, Unit-3 and Unit-2. The data on WFA/LTA elongation, FR growth and bow, WFA bow and twist, RCCA drag force and drag work, RCCA drop time, FR cladding integrity as well as the visual observation of fuel assemblies obtained during the 2006-2012 outages was utilized. The analysis of the measured data showed that assembly growth, FR bow, irradiation growth, and Zr-1%Nb grid and ZIRLO cladding corrosion lies within the design limits. The RCCA drop time measured for the LTA/WFA is about 1.9 s at BOC and practically does not change at EOC. The measured WFA bow and twist, and data of drag work on RCCA insertion showed that the WFA deformation in the mixed core is mostly controlled by the distortion of Russian FAs (TVSA) having the higher lateral stiffness. The visual inspection of WFAs carried out during the 2012 outages revealed some damage to the Zr-1%Nb grid outer strap for some WFAs during the loading sequence. The performed fundamental investigations allowed identifying the root cause of grid outer strap deformation and proposing the WFA design modifications for preventing damage to SG at a 225 kg handling trip limit.

  6. Experimental Study of Turbine Fuel Thermal Stability in an Aircraft Fuel System Simulator

    NASA Technical Reports Server (NTRS)

    Vranos, A.; Marteney, P. J.

    1980-01-01

    The thermal stability of aircraft gas turbines fuels was investigated. The objectives were: (1) to design and build an aircraft fuel system simulator; (2) to establish criteria for quantitative assessment of fuel thermal degradation; and (3) to measure the thermal degradation of Jet A and an alternative fuel. Accordingly, an aircraft fuel system simulator was built and the coking tendencies of Jet A and a model alternative fuel (No. 2 heating oil) were measured over a range of temperatures, pressures, flows, and fuel inlet conditions.

  7. Monte Carlo Modeling of Fast Sub-critical Assembly with MOX Fuel for Research of Accelerator-Driven Systems

    NASA Astrophysics Data System (ADS)

    Polanski, A.; Barashenkov, V.; Puzynin, I.; Rakhno, I.; Sissakian, A.

    It is considered a sub-critical assembly driven with existing 660 MeV JINR proton accelerator. The assembly consists of a central cylindrical lead target surrounded with a mixed-oxide (MOX) fuel (PuO2 + UO2) and with reflector made of beryllium. Dependence of the energetic gain on the proton energy, the neutron multiplication coefficient, and the neutron energetic spectra have been calculated. It is shown that for subcritical assembly with a mixed-oxide (MOX) BN-600 fuel (28%PuO 2 + 72%UO2) with effective density of fuel material equal to 9 g/cm 3 , the multiplication coefficient keff is equal to 0.945, the energetic gain is equal to 27, and the neutron flux density is 1012 cm˜2 s˜x for the protons with energy of 660 MeV and accelerator beam current of 1 uA.

  8. Experimental evaluation of combustor concepts for burning broad property fuels

    NASA Technical Reports Server (NTRS)

    Kasper, J. M.; Ekstedt, E. E.; Dodds, W. J.; Shayeson, M. W.

    1980-01-01

    A baseline CF6-50 combustor and three advanced combustor designs were evaluated to determine the effects of combustor design on operational characteristics using broad property fuels. Three fuels were used in each test: Jet A, a broad property 13% hydrogen fuel, and a 12% hydrogen fuel blend. Testing was performed in a sector rig at true cruise and simulated takeoff conditions for the CF6-50 engine cycle. The advanced combustors (all double annular, lean dome designs) generally exhibited lower metal temperatures, exhaust emissions, and carbon buildup than the baseline CF6-50 combustor. The sensitivities of emissions and metal temperatures to fuel hydrogen content were also generally lower for the advanced designs. The most promising advanced design used premixing tubes in the main stage. This design was chosen for additional testing in which fuel/air ratio, reference velocity, and fuel flow split were varied.

  9. Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

    SciTech Connect

    Cuta, Judith M.; Jenquin, Urban P.; McKinnon, Mikal A.

    2001-11-20

    The licensing of spent fuel storage casks is generally based on conservative analyses that assume a storage system being uniformly loaded with design basis fuel. The design basis fuel typically assumes a maximum assembly enrichment, maximum burn up, and minimum cooling time. These conditions set the maximum decay heat loads and radioactive source terms for the design. Recognizing that reactor spent fuel pools hold spent fuel with an array of initial enrichments, burners, and cooling times, this study was performed to evaluate the effect of load pattern on peak cladding temperature and cask surface dose rate. Based on the analysis, the authors concluded that load patterns could be used to reduce peak cladding temperatures in a cask without adversely impacting the surface dose rates.

  10. CFD Analysis of Coolant Flow in VVER-440 Fuel Assemblies with the Code ANSYS CFX 10.0

    SciTech Connect

    Toth, Sandor; Legradi, Gabor; Aszodi, Attila

    2006-07-01

    From the aspect of planning the power upgrading of nuclear reactors - including the VVER-440 type reactor - it is essential to get to know the flow field in the fuel assembly. For this purpose we have developed models of the fuel assembly of the VVER-440 reactor using the ANSYS CFX 10.0 CFD code. At first a 240 mm long part of a 60 degrees segment of the fuel pin bundle was modelled. Implementing this model a sensitivity study on the appropriate meshing was performed. Based on the development of the above described model, further models were developed: a 960 mm long part of a 60-degree-segment and a full length part (2420 mm) of the fuel pin bundle segment. The calculations were run using constant coolant properties and several turbulence models. The impacts of choosing different turbulence models were investigated. The results of the above-mentioned investigations are presented in this paper. (authors)

  11. Field test and evaluation of the IAEA coincidence collar for the measurement of unirradiated BWR fuel assemblies

    SciTech Connect

    Menlove, H.O.; Keddar, A.

    1982-12-01

    The neutron coincidence counter has been field tested and evaluated for the measurement of boiling-water-reactor (BWR) fuel assemblies at the ASEA-ATOM Fuel Fabrication Facility. The system measures the /sup 235/U content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The /sup 238/U content is measured in the passive mode without the AmLi neutron interrogatioin source. The field tests included both standard production movable fuel rods to investigate enrichment and absorber variations. Results gave a response standard deviation of 0.9% for the active case and 2.1% for the passive case in 1000-s measurement times. 10 figures, 2 tables.

  12. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    SciTech Connect

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  13. Experimental Studies of Coal and Biomass Fuel Synthesis and Flame Characterization for Aircraft Engines (Year Two)

    DTIC Science & Technology

    2011-03-31

    2.1 Experimental Investigation of Coal and Biomass Gasification using In-situ Diagnostics ................ 31  2.2 References...need for fundamental scientific and synergistic research in catalytic biomass fast-hydropyrolysis, advanced coal gasification and liquid fuel...experimental findings will improve the scientific knowledge of catalytic biomass fast-hydropyrolysis, coal/ biomass gasification and liquid fuel combustion

  14. An experimental study on thermal stability of biodiesel fuel

    NASA Astrophysics Data System (ADS)

    Zhu, Yiying

    Biodiesel fuel, as renewable energy, has been used in conventional diesel engines in pure form or as biodiesel/diesel blends for many years. However, thermal stability of biodiesel and biodiesel/diesel blends has been minimally explored. Aimed to shorten this gap, thermal stability of biodiesel is investigated at high temperatures. In this study, batch thermal stressing experiments of biodiesel fuel were performed in stainless steel coils at specific temperature and residence time range from 250 to 425 °C and 3 to 63 minutes, respectively. Evidence of different pathways of biodiesel fuel degradation is demonstrated chromatographically. It was found that biodiesel was stable at 275 °C for a residence time of 8 minutes or below, but the cis-trans isomerization reaction was observed at 28 minutes. Along with isomerization, polymerization also took place at 300 °C at 63 minutes. Small molecular weight products were detected at 350 °C at 33 minutes resulting from pyrolysis reactions and at 360 °C for 33 minutes or above, gaseous products were produced. The formed isomers and dimers were not stable, further decomposition of these compounds was observed at high temperatures. These three main reactions and the temperature ranges in which they occurred are: isomerization, 275--400 °C; polymerization (Diels-Alder reaction), 300--425 °C; pyrolysis reaction, ≥350 °C. The longer residence time and higher temperature resulted in greater decomposition. As the temperature increased to 425 °C, the colorless biodiesel became brownish. After 8 minutes, almost 84% of the original fatty acid methyl esters (FAMEs) disappeared, indicating significant fuel decomposition. A kinetic study was also carried out subsequently to gain better insight into the biodiesel thermal decomposition. A three-lump model was proposed to describe the decomposition mechanism. Based on this mechanism, a reversible first-order reaction kinetic model for the global biodiesel decomposition was shown to

  15. Differential die-away instrument: Report on comparison of fuel assembly experiments and simulations

    SciTech Connect

    Goodsell, Alison Victoria; Henzl, Vladimir; Swinhoe, Martyn Thomas; Rael, Carlos D.; Desimone, David J.

    2015-01-14

    Experimental results of the assay of mock-up (fresh) fuel with the differential die-away (DDA) instrument were compared to the Monte Carlo N-Particle eXtended (MCNPX) simulation results. Most principal experimental observables, the die-away time and the in tegral of the DDA signal in several time domains, have been found in good agreement with the MCNPX simulation results. The remaining discrepancies between the simulation and experimental results are likely due to small differences between the actual experimental setup and the simulated geometry, including uncertainty in the DT neutron generator yield. Within this report we also present a sensitivity study of the DDA instrument which is a complex and sensitive system and demonstrate to what degree it can be impacted by geometry, material composition, and electronics performance.

  16. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    SciTech Connect

    Trellue, Holly Renee; Fugate, Michael Lynn; Tobin, Stephen Joesph

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  17. Estimation of Critical Flow Velocity for Collapse of Gas Test Loop Booster Fuel Assembly

    SciTech Connect

    Guillen; Mark J. Russell

    2006-07-01

    This paper presents calculations performed to determine the critical flow velocity for plate collapse due to static instability for the Gas Test Loop booster fuel assembly. Long, slender plates arranged in a parallel configuration can experience static divergence and collapse at sufficiently high coolant flow rates. Such collapse was exhibited by the Oak Ridge High Flux Reactor in the 1940s and the Engineering Test Reactor at the Idaho National Laboratory in the 1950s. Theoretical formulas outlined by Miller, based upon wide-beam theory and Bernoulli’s equation, were used for the analysis. Calculations based upon Miller’s theory show that the actual coolant flow velocity is only 6% of the predicted critical flow velocity. Since there is a considerable margin between the theoretically predicted plate collapse velocity and the design velocity, the phenomena of plate collapse due to static instability is unlikely.

  18. Assessment of the impacts of spent fuel disassembly alternatives on the Nuclear Waste Isolation System. [Preparing and packaging spent fuel assemblies for geologic disposal

    SciTech Connect

    Not Available

    1984-07-01

    The objective of this report was to evaluate four possible alternative methods of preparing and packaging spent fuel assemblies for geologic disposal against the Reference Process of unmodified spent fuel. The four alternative processes were: (1) End fitting removal, (2) Fission gas venting and resealing, (3) Fuel bundle disassembly and close packing of fuel pins, and (4) Fuel shearing and immobilization. Systems analysis was used to develop a basis of comparison of the alternatives. Conceptual processes and facility layouts were devised for each of the alternatives, based on technology deemed feasible for the purpose. Assessments were made of 15 principal attributes from the technical, operational, safety/risk, and economic considerations related to each of the alternatives, including both the surface packaging and underground repository operations. Specific attributes of the alternative processes were evaluated by assigning a number for each that expressed its merit relative to the corresponding attribute of the Reference Process. Each alternative process was then ranked by summing the numbers for attributes in each of the four assessment areas and collectively. Fuel bundle disassembly and close packing of fuel pins was ranked the preferred method of disposal of spent fuel. 63 references, 46 figures, 46 tables.

  19. Development and experimental characterization of a fuel cell powered aircraft

    NASA Astrophysics Data System (ADS)

    Bradley, Thomas H.; Moffitt, Blake A.; Mavris, Dimitri N.; Parekh, David E.

    This paper describes the characteristics and performance of a fuel cell powered unmanned aircraft. The aircraft is novel as it is the largest compressed hydrogen fuel cell powered airplane built to date and is currently the only fuel cell aircraft whose design and test results are in the public domain. The aircraft features a 500 W polymer electrolyte membrane fuel cell with full balance of plant and compressed hydrogen storage incorporated into a custom airframe. Details regarding the design requirements, implementation and control of the aircraft are presented for each major aircraft system. The performances of the aircraft and powerplant are analyzed using data from flights and laboratory tests. The efficiency and component power consumption of the fuel cell propulsion system are measured at a variety of flight conditions. The performance of the aircraft powerplant is compared to other 0.5-1 kW-scale fuel cell powerplants in the literature and means of performance improvement for this aircraft are proposed. This work represents one of the first studies of fuel cell powered aircraft to result in a demonstration aircraft. As such, the results of this study are of practical interest to fuel cell powerplant and aircraft designers.

  20. Effects of fuel nozzle design on performance of an experimental annular combustor using natural gas fuel

    NASA Technical Reports Server (NTRS)

    Wear, J. D.; Schultz, D. F.

    1972-01-01

    Tests of various fuel nozzles were conducted with natural gas fuel in a full-annulus combustor. The nozzles were designed to provide either axial, angled, or radial fuel injection. Each fuel nozzle was evaluated by measuring combustion efficiency at relatively severe combustor operating conditions. Combustor blowout and altitude ignition tests were also used to evaluate nozzle designs. Results indicate that angled injection gave higher combustion efficiency, less tendency toward combustion instability, and altitude relight characteristics equal to or superior to those of the other fuel nozzles that were tested.

  1. Radionuclide characterization of reactor decommissioning waste and spent fuel assembly hardware

    SciTech Connect

    Robertson, D.E.; Thomas, C.W.; Wynhoff, N.L.; Hetzer, D.C. )

    1991-01-01

    This study is providing the NRC and licensees with a more comprehensive and defensible data base and regulatory assessment of the radiological factors associated with reactor decommissioning and disposal of wastes generated during these activities. The objectives of this study are being accomplished during a two-phase sampling, measurement, and assessment program involving the actual decommissioning of Shippingport Station and the detailed analysis of neutron-activated materials from commercial reactors. Radiological characterization studies at Shippingport have shown that neutron activation products, dominated by {sup 60}Co, comprised the residual radionuclide inventory. Fission products and transuranic radionuclides were essentially absent. Waste classification assessments have shown that all decommissioning materials (except reactor pressure vessel internals) could be disposed of as Class A waste. Measurements and assessments of spent fuel assembly hardware have shown that {sup 63}Ni, {sup 59}Ni, and {sup 94}Nb sometimes greatly exceed the 10CFR61 Class C limit for some components, and thus would require disposal in a high level waste repository. These measurements are providing the basis for an assessment of the disposal options for these types of highly radioactive materials. Comparisons of predicted (calculated) activation product concentrations with the empirical data are providing as assessment of the accuracy of calculational methods. Work is continuing on radiological characterization of spent PWR and BWR control rod assemblies. Additional work is planned on current issues/problems relating to reactor decommissioning. These efforts will be reported on in future supplements to this report. 20 refs., 23 figs., 34 tabs.

  2. Quantity Distance for the Kennedy Space Center Vehicle Assembly Building for Solid Propellant Fueled Launchers

    NASA Technical Reports Server (NTRS)

    Stover, Steven; Diebler, Corey; Frazier, Wayne

    2006-01-01

    The NASA KSC VAB was built to process Apollo launchers in the 1960's, and later adapted to process Space Shuttles. The VAB has served as a place to assemble solid rocket motors (5RM) and mate them to the vehicle's external fuel tank and Orbiter before rollout to the launch pad. As Space Shuttle is phased out, and new launchers are developed, the VAB may again be adapted to process these new launchers. Current launch vehicle designs call for continued and perhaps increased use of SRM segments; hence, the safe separation distances are in the process of being re-calculated. Cognizant NASA personnel and the solid rocket contractor have revisited the above VAB QD considerations and suggest that it may be revised to allow a greater number of motor segments within the VAB. This revision assumes that an inadvertent ignition of one SRM stack in its High Bay need not cause immediate and complete involvement of boosters that are part of a vehicle in adjacent High Bay. To support this assumption, NASA and contractor personnel proposed a strawman test approach for obtaining subscale data that may be used to develop phenomenological insight and to develop confidence in an analysis model for later use on full-scale situations. A team of subject matter experts in safety and siting of propellants and explosives were assembled to review the subscale test approach and provide options to NASA. Upon deliberations regarding the various options, the team arrived at some preliminary recommendations for NASA.

  3. Towards neat methanol operation of direct methanol fuel cells: a novel self-assembled proton exchange membrane.

    PubMed

    Li, Jing; Cai, Weiwei; Ma, Liying; Zhang, Yunfeng; Chen, Zhangxian; Cheng, Hansong

    2015-04-18

    We report here a novel proton exchange membrane with remarkably high methanol-permeation resistivity and excellent proton conductivity enabled by carefully designed self-assembled ionic conductive channels. A direct methanol fuel cell utilizing the membrane performs well with a 20 M methanol solution, very close to the concentration of neat methanol.

  4. Artificial Neural Network-Based Monitoring of the Fuel Assembly Temperature Sensor and FPGA Implementation

    SciTech Connect

    2015-07-01

    Numerous methods have been developed around the world to model the dynamic behavior and detect a faulty operating mode of a temperature sensor. In this context, we present in this study a new method based on the dependence between the fuel assembly temperature profile on control rods positions, and the coolant flow rate in a nuclear reactor. This seems to be possible since the insertion of control rods at different axial positions and variations in flow rate of the reactor coolant results in different produced thermal power in the reactor. This is closely linked to the instant fuel rod temperature profile. In a first step, we selected parameters to be used and confirmed the adequate correlation between the chosen parameters and those to be estimated by the proposed monitoring system. In the next step, we acquired and de-noised the data of corresponding parameters, the qualified data is then used to design and train the artificial neural network. The effective data denoising was done by using the wavelet transform to remove a various kind of artifacts such as inherent noise. With the suitable choice of wavelet level and smoothing method, it was possible for us to remove all the non-required artifacts with a view to verify and analyze the considered signal. In our work, several potential mother wavelet functions (Haar, Daubechies, Bi-orthogonal, Reverse Bi-orthogonal, Discrete Meyer and Symlets) were investigated to find the most similar function with the being processed signals. To implement the proposed monitoring system for the fuel rod temperature sensor (03 wire RTD sensor), we used the Bayesian artificial neural network 'BNN' technique to model the dynamic behavior of the considered sensor, the system correlate the estimated values with the measured for the concretization of the proposed system we propose an FPGA (field programmable gate array) implementation. The monitoring system use the correlation. (authors)

  5. Examination of stainless steel-clad Connecticut Yankee fuel assembly S004 after storage in borated water

    SciTech Connect

    Langstaff, D.C.; Bailey, W.J.; Johnson, A.B. Jr.; Landow, M.P.; Pasupathi, V.; Klingensmith, R.W.

    1982-09-01

    A Connecticut Yankee fuel assembly (S004) was tested nondestructively and destructively. It was concluded that no obvious degradation of the 304L stainless steel-clad spent fuel from assembly S004 occurred during 5 y of storage in borated water. Furthermore, no obvious degradation due to the pool environment occurred on 304 stainless steel-clad rods in assemblies H07 and G11, which were stored for shorter periods but contained operationally induced cladding defects. The seam welds in the cladding on fuel rods from assembly S004, H07, and G11 were similar in that they showed a wrought microstructure with grains noticeably smaller than those in the cladding base metal. The end cap welds showed a dendritically cored structure, typical of rapidly quenched austenitic weld metal. Some intergranular melting may have occurred in the heat-affected zone (HAZ) in the cladding adjacent to the end cap welds in rods from assemblies S004 and H07. However, the weld areas did not show evidence of corrosion-induced degradation.

  6. Original Experimental Approach for Assessing Transport Fuel Stability

    PubMed Central

    Bacha, Kenza; Ben Amara, Arij; Alves Fortunato, Maira; Wund, Perrine; Veyrat, Benjamin; Hayrault, Pascal; Vannier, Axel; Nardin, Michel; Starck, Laurie

    2016-01-01

    The study of fuel oxidation stability is an important issue for the development of future fuels. Diesel and kerosene fuel systems have undergone several technological changes to fulfill environmental and economic requirements. These developments have resulted in increasingly severe operating conditions whose suitability for conventional and alternative fuels needs to be addressed. For example, fatty acid methyl esters (FAMEs) introduced as biodiesel are more prone to oxidation and may lead to deposit formation. Although several methods exist to evaluate fuel stability (induction period, peroxides, acids, and insolubles), no technique allows one to monitor the real-time oxidation mechanism and to measure the formation of oxidation intermediates that may lead to deposit formation. In this article, we developed an advanced oxidation procedure (AOP) based on two existing reactors. This procedure allows the simulation of different oxidation conditions and the monitoring of the oxidation progress by the means of macroscopic parameters, such as total acid number (TAN) and advanced analytical methods like gas chromatography coupled to mass spectrometry (GC-MS) and Fourier Transform Infrared - Attenuated Total Reflection (FTIR-ATR). We successfully applied AOP to gain an in-depth understanding of the oxidation kinetics of a model molecule (methyl oleate) and commercial diesel and biodiesel fuels. These developments represent a key strategy for fuel quality monitoring during logistics and on-board utilization. PMID:27805606

  7. Experimentally testing and assessing the predictive power of species assembly rules for tropical canopy ants

    PubMed Central

    Fayle, Tom M; Eggleton, Paul; Manica, Andrea; Yusah, Kalsum M; Foster, William A

    2015-01-01

    Understanding how species assemble into communities is a key goal in ecology. However, assembly rules are rarely tested experimentally, and their ability to shape real communities is poorly known. We surveyed a diverse community of epiphyte-dwelling ants and found that similar-sized species co-occurred less often than expected. Laboratory experiments demonstrated that invasion was discouraged by the presence of similarly sized resident species. The size difference for which invasion was less likely was the same as that for which wild species exhibited reduced co-occurrence. Finally we explored whether our experimentally derived assembly rules could simulate realistic communities. Communities simulated using size-based species assembly exhibited diversities closer to wild communities than those simulated using size-independent assembly, with results being sensitive to the combination of rules employed. Hence, species segregation in the wild can be driven by competitive species assembly, and this process is sufficient to generate observed species abundance distributions for tropical epiphyte-dwelling ants. PMID:25622647

  8. Improved strategies for fuel assembly, pin cell and reflector cross section generation using the discrete ordinates code DORT

    SciTech Connect

    Pautz, A.

    2006-07-01

    Additional functionality has been added to the Discrete Ordinates transport code DORT in order to produce few-group, homogenized cross sections for typical fuel assembly geometries, both on the assembly and the pin cell level. It is demonstrated, that even on the pin-by-pin level almost perfect reaction rate and pin power conservation can be achieved by using the so called Super-homogenization (SPH) algorithm. This method also allows the generation of appropriate reflector cross sections, which can significantly improve the quality of pin power values in the vicinity of moderator regions. The effectiveness of this approach is demonstrated on several examples, including single fuel assembly calculations as well as the C5G7-MOX and the recent NBA VENUS-7 plutonium recycling benchmark problems. (authors)

  9. In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

    SciTech Connect

    Swinhoe, Martyn Thomas; De Baere, Paul

    2015-04-17

    A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hour of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.

  10. Co-flow anode/cathode supply heat exchanger for a solid-oxide fuel cell assembly

    DOEpatents

    Haltiner, Jr., Karl J.; Kelly, Sean M.

    2005-11-22

    In a solid-oxide fuel cell assembly, a co-flow heat exchanger is provided in the flow paths of the reformate gas and the cathode air ahead of the fuel cell stack, the reformate gas being on one side of the exchanger and the cathode air being on the other. The reformate gas is at a substantially higher temperature than is desired in the stack, and the cathode gas is substantially cooler than desired. In the co-flow heat exchanger, the temperatures of the reformate and cathode streams converge to nearly the same temperature at the outlet of the exchanger. Preferably, the heat exchanger is formed within an integrated component manifold (ICM) for a solid-oxide fuel cell assembly.

  11. Fuel Canister Stress Corrosion Cracking Susceptibility Experimental Results

    SciTech Connect

    Colleen Shelton-Davis

    2003-03-01

    The National Spent Nuclear Fuel Program is tasked with ensuring the U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) is acceptable for permanent disposal at a designated repository. From a repository acceptance criteria viewpoint and from a transportation viewpoint, of significant concern is the condition of the container at the time of shipment. Because the fuel will be in temporary storage for as much as 50 years, verification that no significant degradation has occurred to the canister is required to preclude repackaging all the fuel. Many canisters are also being removed from wet storage, vacuum dried (hot or cold), and then placed into dry storage. This process could have a detrimental effect on canister integrity. Research is currently underway to provide a technically sound assessment of the expected canister condition at the end of interim storage.

  12. Experimental and Modeling Studies of the Combustion Characteristics of Conventional and Alternative Jet Fuels. Final Report

    NASA Technical Reports Server (NTRS)

    Meeks, Ellen; Naik, Chitral V.; Puduppakkam, Karthik V.; Modak, Abhijit; Egolfopoulos, Fokion N.; Tsotsis, Theo; Westbrook, Charles K.

    2011-01-01

    The objectives of this project have been to develop a comprehensive set of fundamental data regarding the combustion behavior of jet fuels and appropriately associated model fuels. Based on the fundamental study results, an auxiliary objective was to identify differentiating characteristics of molecular fuel components that can be used to explain different fuel behavior and that may ultimately be used in the planning and design of optimal fuel-production processes. The fuels studied in this project were Fischer-Tropsch (F-T) fuels and biomass-derived jet fuels that meet certain specifications of currently used jet propulsion applications. Prior to this project, there were no systematic experimental flame data available for such fuels. One of the key goals has been to generate such data, and to use this data in developing and verifying effective kinetic models. The models have then been reduced through automated means to enable multidimensional simulation of the combustion characteristics of such fuels in real combustors. Such reliable kinetic models, validated against fundamental data derived from laminar flames using idealized flow models, are key to the development and design of optimal combustors and fuels. The models provide direct information about the relative contribution of different molecular constituents to the fuel performance and can be used to assess both combustion and emissions characteristics.

  13. Experimental and Analytical Study of Balanced-Diaphragm Fuel Distributors for Gas-Turbine Engines

    NASA Technical Reports Server (NTRS)

    Straight, David M.; Gold, Harold

    1950-01-01

    A method of distributing fuel equally to a plurality of spray nozzles in a gas-turbine engine by means of balanced-diaphragm fuel distributors is presented. The experimental performance of three of eight possible distributor arrangements are discussed. An analysis of all eight arrangements is included. Criterions are given for choosing a fuel-distributor arrangement to meet specific fuel-system requirements of fuel-distribution accuracy, spray-nozzle pressure variations, and fuel-system pressures. Data obtained with a model of one distributor arrangement indicated a maximum deviation from perfect distribution of 3.3 percent for a 44 to 1 range (19.5 to 862 lb/hr) of fuel-flow rates. The maximum distributor pressure drop was 125 pounds per square inch. The method used to obtain the required wide range of flow control in the distributor valves consisted in varying the length of a constant-area flow path.

  14. Fuel-optimal slewing of an experimental hinged-free beam

    NASA Technical Reports Server (NTRS)

    Silverberg, Larry; Meyer, John L.

    1993-01-01

    With the recent development of numerical methods for exactly solving larger order fuel-optlmal control problems, the implementation of exact real-time fuel-optimal control becomes possible. This paper describes a physical experiment in which a hinged-free beam is slewed in a fuel-optimal manner. A comparison between experimental results and the analytical predictions highlights the sources of errors typical of this class of problems.

  15. Experimental Study of the Stability of Aircraft Fuels at Elevated Temperatures

    NASA Technical Reports Server (NTRS)

    Vranos, A.; Marteney, P. J.

    1980-01-01

    An experimental study of fuel stability was conducted in an apparatus which simulated an aircraft gas turbine fuel system. Two fuels were tested: Jet A and Number 2 Home Heating oil. Jet A is an aircraft gas turbine fuel currently in wide use. No. 2HH was selected to represent the properties of future turbine fuels, particularly experimental Reference Broad Specification, which, under NASA sponsorship, was considered as a possible next-generation fuel. Tests were conducted with varying fuel flow rates, delivery pressures and fuel pretreatments (including preheating and deoxygenation). Simulator wall temperatures were varied between 422K and 672K at fuel flows of 0.022 to 0.22 Kg/sec. Coking rate was determined at four equally-spaced locations along the length of the simulator. Fuel samples were collected for infrared analysis. The dependence of coking rate in Jet A may be correlated with surface temperature via an activation energy of 9 to 10 kcal/mole, although the results indicate that both bulk fluid and surface temperature affect the rate of decomposition. As a consequence, flow rate, which controls bulk temperature, must also be considered. Taken together, these results suggest that the decomposition reactions are initiated on the surface and continue in the bulk fluid. The coking rate data for No. 2 HH oil are very highly temperature dependent above approximately 533K. This suggests that bulk phase reactions can become controlling in the formation of coke.

  16. Experimental study of the thermal stability of hydrocarbon fuels

    NASA Technical Reports Server (NTRS)

    Marteney, P. J.; Colket, M. B.; Vranos, A.

    1982-01-01

    The thermal stability of two hydrocarbon fuels (premium diesel and regular diesel) was determined in a flow reactor under conditions representing operation of an aircraft gas turbine engine. Temperature was varied from 300 to 750 F (422 to 672 K) for fuel flows of 2.84 to 56.8 liters/hr (corresponding to 6.84 x 0.00010 to 1.63 x 0.010 kg/sec for regular diesel fuel and 6.55 x 0.00010 to 1.37 x 0.010 kg/sec for premium diesel fuel); test times varied between 1 and 8 hr. The rate of deposition was obtained through measurement of weight gained by metal discs fixed along the channel wall. The rate of deposit formation is best correlated by an Arrhenius expression. The sample discs in the flow reactor were varied among stainless steel, aluminum and brass; fuels were doped with quinoline, indole, and benzoyl perioxide to yield nitrogen or oxygen concentrations of approximately 1000 ppm. The most substantial change in rate was an increase in deposits for brass discs; other disc materials or the additives caused only small perturbations. Tests were also conducted in a static reactor at temperatures of 300 to 800 F for times of 30 min to 2 1/2 hr. Much smaller deposition was found, indicating the importance of fluid transport in the mechanism.

  17. Assessment of the integrity of spent fuel assemblies used in dry storage demonstrations at the Nevada Test Site

    SciTech Connect

    Johnson, A.B. Jr.; Dobbins, J.C.; Zaloudek, F.R.

    1987-07-01

    This report summarizes the histories of 17 Zircaloy-clad spent fuel assemblies used in dry storage tests and demonstrations at the Engine Maintenance and Disassembly (EMAD) and Climax facilities at the Nevada Test Site (NTS). The 18th assembly was shipped to the Battelle Columbus Laboratory (BCL) and remained there for extensive characterization and as a source of specimens for whole-rod and rod-segment dry storage tests. The report traces the history of the assemblies after discharge from the Turkey Point Unit 3 pressurized-water reactor (1975 and 1977) through shipment (first arrival at EMAD in December 1978), dry storage tests and demonstrations, and shipment by truck cask from EMAD to the Idaho National Engineering Laboratory (INEL) in May/June 1986. The principal objectives of this report are to assess and document the integrity of the fuel during the extensive dry storage activities at NTS and BCL, and to briefly summarize the dry storage technologies and procedures demonstrated in this program. The dry storage tests and demonstrations involved the following concepts and facilities: (1) surface drywells (EMAD); (2) deep drywells (425 m underground in the Climax granite formation); (3) concrete silo (EMAD); (4) air-cooled vault (EMAD); (5) electrically-heated module for fuel assembly thermal calibration and testing (EMAD/FAITM). 20 refs., 43 figs., 9 tabs.

  18. Improvement of the thermal margins in the Swedish Ringhals-3 PWR by introducing new fuel assemblies with thorium

    SciTech Connect

    Lau, C. W.; Demaziere, C.; Nylen, H.; Sandberg, U.

    2012-07-01

    Thorium is a fertile material and most of the past research has focused on breeding thorium to fissile material. In this paper, the focus is on using thorium to improve the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. All the key safety parameters, such as isothermal temperature coefficient of reactivity, Doppler temperature of reactivity, boron worth, shutdown margins and fraction of delayed neutrons are studied in this paper, and are within safety limits for the new core design using the uranium-thorium-based fuel assemblies. The calculations were performed by the two-dimensional transport code CASMO-4E and the two group steady-state three dimensional nodal code SIMULATE-3 from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core designs with less neutron leakage or could be used in power uprates to offer efficient safety margins. (authors)

  19. Determination of optimal imaging parameters for the reconstruction of a nuclear fuel assembly using limited angle neutron tomography

    NASA Astrophysics Data System (ADS)

    Abir, M. I.; Islam, F. F.; Craft, A.; Williams, W. J.; Wachs, D. M.; Chichester, D. L.; Meyer, M. K.; Lee, H. K.

    2016-01-01

    The core components of nuclear reactors (e.g., fuel assemblies, spacer grids, control rods) encounter harsh environments due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of nuclear power plants; post-irradiation examination (PIE) can reveal information about the integrity of these components. Neutron computed tomography (CT) is one important PIE measurement tool for nondestructively evaluating the structural integrity of these items. CT typically requires many projections to be acquired from different view angles, after which a mathematical algorithm is used for image reconstruction. However, when working with heavily irradiated materials and irradiated nuclear fuel, obtaining many projections is laborious and expensive. Image reconstruction from a smaller number of projections has been explored to achieve faster and more cost-efficient PIE. Classical reconstruction methods (e.g., filtered backprojection), unfortunately, do not typically offer stable reconstructions from a highly asymmetric, few-projection data set and often create severe streaking artifacts. We propose an iterative reconstruction technique to reconstruct curved, plate-type nuclear fuel assemblies using limited-angle CT. The performance of the proposed method is assessed using simulated data and validated through real projections. We also discuss the systematic strategy for establishing the conditions of reconstructions and finding the optimal imaging parameters for reconstructions of the fuel assemblies from few projections using limited-angle CT. Results show that a fuel assembly can be reconstructed using limited-angle CT if 36 or more projections are taken from a particular direction with 1° angular increment.

  20. Membrane-less cloth cathode assembly (CCA) for scalable microbial fuel cells.

    PubMed

    Zhuang, Li; Zhou, Shungui; Wang, Yueqiang; Liu, Chengshuai; Geng, Shu

    2009-08-15

    One of the main challenges for scaling up microbial fuel cell (MFC) technologies is developing low-cost cathode architectures that can generate high power output. This study developed a simple method to convert non-conductive material (canvas cloth) into an electrically conductive and catalytically active cloth cathode assembly (CCA) in one step. The membrane-less CCA was simply constructed by coating the cloth with conductive paint (nickel-based or graphite-based) and non-precious metal catalyst (MnO(2)). Under the fed-batch mode, the tubular air-chamber MFCs equipped with Ni-CCA and graphite-CCA generated the maximum power densities of 86.03 and 24.67 mW m(-2) (normalized to the projected cathode surface area), or 9.87 and 2.83 W m(-3) (normalized to the reactor liquid volume), respectively. The higher power output of Ni-CCA-MFC was associated with the lower volume resistivity of Ni-CCA (1.35 x 10(-2)Omega cm) than that of graphite-CCA (225 x 10(-2)Omega cm). At an external resistance of 100 Omega, Ni-CCA-MFC and graphite-CCA-MFC removed approximately 95% COD in brewery wastewater within 13 and 18d, and achieved coulombic efficiencies of 30.2% and 19.5%, respectively. The accumulated net water loss through the cloth by electro-osmotic drag exhibited a linear correlation (R(2)=0.999) with produced coulombs. With a comparable power production, such CCAs only cost less than 5% of the previously reported membrane cathode assembly. The new cathode configuration here is a mechanically durable, economical system for MFC scalability.

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

    NASA Technical Reports Server (NTRS)

    Stockemer, Francis J.

    1979-01-01

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

  2. Decay characteristics of once-through LWR and LMFBR spent fuels, high-level wastes, and fuel-assembly structural material wastes

    SciTech Connect

    Croff, A.G.; Alexander, C.W.

    1980-11-01

    The decay characteristics of spent fuel, high-level waste, and fuel-assembly structural material (cladding) waste are presented in the form of ORIGEN2 output tables for (1) a pressurized water reactor operating on a once-through cycle with low-enrichment uranium feed, (2) a boiling-water reactor operating on a once-through cycle with low-enrichment uranium feed, and (3) a liquid-metal fast breeder reactor being fueled with depleted uranium enriched with discharged light water reactor plutonium on a once-through basis. The decay characteristics given include the mass (g), radioactivity (Ci), thermal power (W), photon activity (photons/s and MeV/W-s in 18 energy groups), and neutron activity (neutrons/s) from (..cap alpha..,n) and spontaneous fission events. The first three characteristics are given for each element and for the principal nuclide contributors to the activation products, actinides, and fission products. Also included are a summary description of the ORIGEN2 reactor models that form the basis for the calculated results and a physical description of the fuel assemblies for the three reactors.

  3. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    SciTech Connect

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

  4. Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

    DOE PAGES

    Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.; ...

    2015-10-09

    Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

  5. Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel-Design concept and experimental demonstration

    NASA Astrophysics Data System (ADS)

    Henzlova, D.; Menlove, H. O.; Rael, C. D.; Trellue, H. R.; Tobin, S. J.; Park, Se-Hwan; Oh, Jong-Myeong; Lee, Seung-Kyu; Ahn, Seong-Kyu; Kwon, In-Chan; Kim, Ho-Dong

    2016-01-01

    This paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. This paper describes the initial feasibility demonstration of the CIPN instrument, which involved measurements of four pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. These features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.

  6. Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

    SciTech Connect

    Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.; Trellue, Holly Renee; Tobin, Stephen Joseph; Park, Se-Hwan; Oh, Jong-Myeong; Lee, Seung-Kyu; Ahn, Seong-Kyu; Kwon, In-Chan; Kim, Ho-Dong

    2015-10-09

    Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of four pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.

  7. Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

    DOEpatents

    Zafred, Paolo R [Murrysville, PA; Gillett, James E [Greensburg, PA

    2012-04-24

    A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

  8. Experimental study of bioelectrochemical fuel cell using bacteria from baltic sea

    NASA Astrophysics Data System (ADS)

    Halme, A.; Zhang, X.

    1995-02-01

    A bioelectrochemical fuel cell using bacteria as catalyst was investigated in this paper. The bacteria were obtained from the bottom sediment of Baltic Sea, and then cultivated in a 1 liter bioreactor. Raw material for fermentation were glucose first and then fish meat or plankton biomass. After certain fermentation time, broth was used as fuel for the fuel cell. A steady power output (200 microW/ml anodic volume) was obtained by using stainless steel net packing with graphite particles as the anode electrode. Different fermentation conditions were tested for maximum electroactive substance output. The experimental study of the fuel cell were carried out as follows: (1) characteristics of the fuel cell; (2) mediator effect on the current output; and (3) mode of the fuel flow.

  9. Electric power generation by a submersible microbial fuel cell equipped with a membrane electrode assembly.

    PubMed

    Min, Booki; Poulsen, Finn Willy; Thygesen, Anders; Angelidaki, Irini

    2012-08-01

    Membrane electrode assemblies (MEAs) were incorporated into the cathode chamber of a submersible microbial fuel cell (SMFC). A close contact of the electrodes could produce high power output from SMFC in which anode and cathode electrodes were connected in parallel. In polarization test, the maximum power density was 631 mW/m(2) at current density of 1772 mA/m(2) at 82 Ω. With 180-Ω external resistance, one set of the electrodes on the same side could generate more power density of 832±4 mW/m(2) with current generation of 1923±4 mA/m(2). The anode, inclusive a biofilm behaved ohmic, whereas a Tafel type behavior was observed for the oxygen reduction. The various impedance contributions from electrodes, electrolyte and membrane were analyzed and identified by electrochemical impedance spectroscopy. Air flow rate to the cathode chamber affected microbial voltage generation, and higher power generation was obtained at relatively low air flow less than 2 mL/min.

  10. Nuclear reactor fuel assembly duct-tube-to-handling-socket attachment system

    DOEpatents

    Christiansen, David W.; Smith, Bob G.

    1982-01-01

    A reusable system for removably attaching the upper end 10of a nuclear reactor duct tube to the lower end 30 of a nuclear reactor fuel assembly handling socket. A transition ring 20, fixed to the duct tube's upper end 10, has an interior-threaded section 22 with a first locking hole segment 24. An adaptor ring 40, fixed to the handling socket's lower end 30 has an outside-threaded section 42 with a second locking hole segment 44. The inside 22 and outside 42 threaded sections match and can be joined so that the first 24 and second 44 locking hole segments can be aligned to form a locking hole. A locking ring 50, with a locking pin 52, slides over the adaptor ring 40 so that the locking pin 52 fits in the locking hole. A swage lock 60 or a cantilever finger lock 70 is formed from the locking cup collar 26 to fit in a matching groove 54 or 56 in the locking ring 50 to prevent the locking ring's locking pin 52 from backing out of the locking hole.

  11. Treating refinery wastewaters in microbial fuel cells using separator electrode assembly or spaced electrode configurations.

    PubMed

    Zhang, Fang; Ahn, Yongtae; Logan, Bruce E

    2014-01-01

    The effectiveness of refinery wastewater (RW) treatment using air-cathode, microbial fuel cells (MFCs) was examined relative to previous tests based on completely anaerobic microbial electrolysis cells (MECs). MFCs were configured with separator electrode assembly (SEA) or spaced electrode (SPA) configurations to measure power production and relative impacts of oxygen crossover on organics removal. The SEA configuration produced a higher maximum power density (280±6 mW/m(2); 16.3±0.4 W/m(3)) than the SPA arrangement (255±2 mW/m(2)) due to lower internal resistance. Power production in both configurations was lower than that obtained with the domestic wastewater (positive control) due to less favorable (more positive) anode potentials, indicating poorer biodegradability of the RW. MFCs with RW achieved up to 84% total COD removal, 73% soluble COD removal and 92% HBOD removal. These removals were higher than those previously obtained in mini-MEC tests, as oxygen crossover from the cathode enhanced degradation in MFCs compared to MECs.

  12. Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

    NASA Astrophysics Data System (ADS)

    Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

    2017-01-01

    The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

  13. First principles Candu fuel model and validation experimentation

    SciTech Connect

    Corcoran, E.C.; Kaye, M.H.; Lewis, B.J.; Thompson, W.T.; Akbari, F.; Higgs, J.D.; Verrall, R.A.; He, Z.; Mouris, J.F.

    2007-07-01

    Many modeling projects on nuclear fuel rest on a quantitative understanding of the co-existing phases at various stages of burnup. Since the various fission products have considerably different abilities to chemically associate with oxygen, and the O/M ratio is slowly changing as well, the chemical potential (generally expressed as an equivalent oxygen partial pressure) is a function of burnup. Concurrently, well-recognized small fractions of new phases such as inert gas, noble metals, zirconates, etc. also develop. To further complicate matters, the dominant UO{sub 2} fuel phase may be non-stoichiometric and most of minor phases have a variable composition dependent on temperature and possible contact with the coolant in the event of a sheathing defect. A Thermodynamic Fuel Model to predict the phases in partially burned Candu nuclear fuel containing many major fission products has been under development. This model is capable of handling non-stoichiometry in the UO{sub 2} fluorite phase, dilute solution behaviour of significant solute oxides, noble metal inclusions, a second metal solid solution U(Pd-Rh-Ru)3, zirconate and uranate solutions as well as other minor solid phases, and volatile gaseous species. The treatment is a melding of several thermodynamic modeling projects dealing with isolated aspects of this important multi-component system. To simplify the computations, the number of elements has been limited to twenty major representative fission products known to appear in spent fuel. The proportion of elements must first be generated using SCALES-5. Oxygen is inferred from the concentration of the other elements. Provision to study the disposition of very minor fission products is included within the general treatment but these are introduced only on an as needed basis for a particular purpose. The building blocks of the model are the standard Gibbs energies of formation of the many possible compounds expressed as a function of temperature. To these data

  14. Development and Experimental Evaluation of Passive Fuel Cell Thermal Control

    NASA Technical Reports Server (NTRS)

    Colozza, Anthony J.; Jakupca, Ian J.; Castle, Charles H.; Burke, Kenneth A.

    2014-01-01

    To provide uniform cooling for a fuel cell stack, a cooling plate concept was evaluated. This concept utilized thin cooling plates to extract heat from the interior of a fuel cell stack and move this heat to a cooling manifold where it can be transferred to an external cooling fluid. The advantages of this cooling approach include a reduced number of ancillary components and the ability to directly utilize an external cooling fluid loop for cooling the fuel cell stack. A number of different types of cooling plates and manifolds were developed. The cooling plates consisted of two main types; a plate based on thermopyrolytic graphite (TPG) and a planar (or flat plate) heat pipe. The plates, along with solid metal control samples, were tested for both thermal and electrical conductivity. To transfer heat from the cooling plates to the cooling fluid, a number of manifold designs utilizing various materials were devised, constructed, and tested. A key aspect of the manifold was that it had to be electrically nonconductive so it would not short out the fuel cell stack during operation. Different manifold and cooling plate configurations were tested in a vacuum chamber to minimize convective heat losses. Cooling plates were placed in the grooves within the manifolds and heated with surface-mounted electric pad heaters. The plate temperature and its thermal distribution were recorded for all tested combinations of manifold cooling flow rates and heater power loads. This testing simulated the performance of the cooling plates and manifold within an operational fuel cell stack. Different types of control valves and control schemes were tested and evaluated based on their ability to maintain a constant temperature of the cooling plates. The control valves regulated the cooling fluid flow through the manifold, thereby controlling the heat flow to the cooling fluid. Through this work, a cooling plate and manifold system was developed that could maintain the cooling plates

  15. Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

    SciTech Connect

    Hofman, G.L.; Neimark, L.A.

    1985-01-01

    Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of /sup 235/U. Fuel plates containing 33 v/o U/sub 3/Si and U/sub 3/Si/sub 2/ behaved very well up to this burnup. Plates containing 33 v/o U/sub 3/Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U/sub 3/Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U/sub 3/Si showed an increase swelling rate around 85% burnup. 5 refs., 10 figs.

  16. ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

    SciTech Connect

    Lell, R. M.; Schaefer, R. W.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Nuclear Engineering Division; Inst. of Physics and Power Engineering

    2007-10-01

    Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of demonstration

  17. In vivo microtubule dynamics during experimentally induced conversions between tubulin assembly states in Allogromia laticollaris.

    PubMed

    Welnhofer, E A; Travis, J L

    1996-01-01

    A distinctive property of foraminiferan tubulin is that, in addition to microtubules (MTs), it exists in an alternate assembly state, helical filaments. Here, we have examined in vivo MT dynamics during experimentally induced conversions between these two assembly states in the reticulopods of the marine foraminiferan Allogromia laticollaris. Exposure to high extracellular concentrations of Mg2+ (165 mM) resulted in a complete conversion of MTs into helical filaments. However, Mg2+ treatment also induced a retrograde movement of organelles and cytoplasm, and it was necessary to inhibit this response in order to assess the effects of assembly state changes on individual MTs. This was accomplished by simultaneous treatment with high extracellular Mg2+ and 2,4-dinitrophenol (DNP). The resulting loss in MTs was detected by video enhanced DIC (VEC-DIC) microscopy as either an endwise MT shortening (at an average rate of 474 microns/min) or transformation into one or more irregularly shaped fibrils, which we termed residual fibrils. Correlative immunofluorescence and video microscopy showed residual fibrils to be composed of helical filaments. Removal of extracellular Mg2+/DNP initiated a reversal in assembly state, from helical filaments into MTs, which was completed within 5 min. VEC-DIC microscopy showed that MTs reformed by an endwise lengthening at an average rate of 216 microns/min. These results suggest that conversion between alternate tubulin assembly states provides a more rapid means to build and dismantle MTs than conventional subunit-driven pathways.

  18. Computational and Experimental Studies of Jet Fuel Combustion

    DTIC Science & Technology

    2009-08-14

    reference flame with a prescribed velocity and thermal field and then to perturb such a flame with known amounts of either jet fuel or surrogates. This... thermal conductivity (TCD), flame ionization (FID) and mass spectrometry detectors (MSD) (Agilent 5973N). The instrument can separate and quantify...location of the peak temperature was approached. This sequence was in line with the anticipated kinetic behavior based on thermal decomposition of

  19. Experimental Determination of Metal Fuel Point Defect Parameters

    SciTech Connect

    Fluss, M J; McCall, S

    2008-06-03

    Nuclear metallic fuels are one of many options for advanced nuclear fuel cycles because they provide dimensional stability, mechanical integrity, thermal efficiency, and irradiation resistance while the associated pyro-processing is technically relevant to concerns about proliferation and diversion of special nuclear materials. In this presentation we will discuss recent success that we have had in studying isochronal annealing of damage cascades in Pu and Pu(Ga) arising from the self-decay of Pu as well as the annealing characteristics of noninteracting point defect populations produced by ion accelerator irradiation. Comparisons of the annealing properties of these two populations of defects arising from very different source terms are enlightening and point to complex defect and mass transport properties in the plutonium specimens which we are only now starting to understand as a result of many follow-on studies. More importantly however, the success of these measurements points the way to obtaining important mass transport parameters for comparison with theoretical predictions or to use directly in existing and future materials modeling of radiation effects in nuclear metallic fuels. The way forward on such measurements and the requisite theory and modeling will be discussed.

  20. HYDRA-I: a three-dimensional finite difference code for calculating the thermohydraulic performance of a fuel assembly contained within a canister

    SciTech Connect

    McCann, R.A.

    1980-12-01

    A finite difference computer code, named HYDRA-I, has been developed to simulate the three-dimensional performance of a spent fuel assembly contained within a cylindrical canister. The code accounts for the coupled heat transfer modes of conduction, convection, and radiation and permits spatially varying boundary conditions, thermophysical properties, and power generation rates. This document is intended as a manual for potential users of HYDRA-I. A brief discussion of the governing equations, the solution technique, and a detailed description of how to set up and execute a problem are presented. HYDRA-I is designed for operation on a CDC 7600 computer. An appendix is included that summarizes approximately two dozen different cases that have been examined. The cases encompass variations in fuel assembly and canister configurations, power generation rates, filler materials, and gases. The results presented show maximum and various local temperatures and heat fluxes illustrating the changing importance of the three heat transfer modes. Finally, the need for comparison with experimental data is emphasized as an aid in code verification although the limited data available indicate excellent agreement.

  1. Hydrocarbon fuel effects in solid-oxide fuel cell operation: an experimental and modeling study of n-hexane pyrolysis.

    PubMed

    Randolph, Katie L; Dean, Anthony M

    2007-08-21

    Pyrolysis experiments of n-hexane were performed and the product distribution and fuel consumption were measured as a function of temperature. The experimental temperatures ranged from 550-675 degrees C, with a pressure of approximately 1 atm, and residence times of approximately 5 s. N-Hexane was used as a model compound to represent the linear alkanes that might be found in practical hydrocarbon fuels. Under these conditions, high fuel conversion was observed at the higher temperatures and a wide range of products were formed. The experimental observations were compared to predictions from a plug-flow model using a reaction mechanism consisting of 205 species and 1403 reactions. The hydrogen abstraction and isomerization rate coefficients in this model were based on CBS-QB3 calculations. The only model modification was adjustment of the A-factor of the initiation rates to match conversion at one temperature. This model was able to successfully predict the observed trends in both product selectivities as well as fuel conversion over the temperature range. The mechanism was also used to capture the trends previously observed in n-butane pyrolysis under similar experimental conditions. Significant differences in the sensitivity coefficients for the hexane and butane systems are discussed in terms of the competition between beta-scission and isomerization of the initial radicals formed. The kinetic model predicts that n-hexane will be completely converted within 0.1 s in the higher temperature environment ( approximately 800 degrees C) of the anode channel of a solid-oxide fuel cell (SOFC). This result clearly illustrates the need to explicitly account for gas-phase reactions in SOFC models for those cases where hydrocarbons, especially those larger than methane, are fed directly to an SOFC.

  2. Effect of placing different obstacles in flow fields on performance of a PEM fuel cell: numerical investigation and experimental comparison

    NASA Astrophysics Data System (ADS)

    Khazaee, I.

    2013-09-01

    In this study a complete two-dimensional model for proton exchange membrane (PEM) fuel cells was used to investigate the effect of using different obstacles on the performances, current density and gas concentration for different aspect ratios (ARs). The proposed model is a full cell model, which includes all the parts of the PEM fuel cell, flow channels, gas diffusion electrodes, catalyst layers and the membrane. Also a series of tests are carried out to investigate and validate the numerical results of the polarization curve under the normal conditions. A PEM fuel cell with 25 cm2 active area and Nafion 117 membrane with 4 mg Pt/cm2 for the anode and cathode is employed as a membrane electrode assembly. The results show that the predicted polarization curves by using this model are in good agreement with the experimental results. Also the results show that the local current density reduces more obviously at a higher overpotential than at a lower overpotential because of the more obvious reflection phenomena in the downstream region. At lower operating voltage conditions, the overall cell performance decreases as the AR decreases.

  3. CFD Simulations of a Flow Mixing and Heat Transfer Enhancement in an Advanced LWR Nuclear Fuel Assembly

    SciTech Connect

    In, Wang-Kee; Chun, Tae-Hyun; Shin, Chang-Hwan; Oh, Dong-Seok

    2007-07-01

    A computational fluid dynamics (CFD) analysis has been performed to investigate a flow-mixing and heat-transfer enhancement caused by a mixing-vane spacer in a LWR fuel assembly which is a rod bundle. This paper presents the CFD simulations of a flow mixing and heat transfer in a fully heated 5x5 array of a rod bundle with a split-vane and hybrid-vane spacer. The CFD prediction at a low Reynolds number of 42,000 showed a reasonably good agreement of the initial heat transfer enhancement with the measured one for a partially heated experiment using a similar spacer structure. The CFD simulation also predicted the decay rate of a normalized Nusselt number downstream of the split-vane spacer which agrees fairly well with those of the experiment and the correlation. The CFD calculations for the split vane and hybrid vane at the LWR operating conditions(Re = 500,000) predicted hot fuel spots in a streaky structure downstream of the spacer, which occurs due to the secondary flow occurring in an opposite direction near the fuel rod. However, the split-vane and hybrid-vane spacers are predicted to significantly enhance the overall heat transfer of a LWR nuclear fuel assembly. (authors)

  4. Development of the JAERI (Japan Atomic Energy Research Institute) fuel cleanup system for tests at the Tritium Systems Test Assembly

    SciTech Connect

    Konishi, S.; Inoue, M.; Hayashi, T.; Okuno, K.; Naruse, Y. ); Barnes, J.W.; Anderson, J.L. )

    1990-01-01

    Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI) has developed the Fuel Cleanup System (FCU) which accepts simulated fusion reactor exhaust and produces pure hydrogen isotopes and tritium-free waste. The major components are: a palladium diffuser, a catalytic reactor, cold traps, a ceramic electrolysis cell, and zirconium-cobalt beds. In 1988, an integrated loop of the FCU process was installed in the TPL and a number of hot'' runs were performed to study the system characteristics and improve system performance. Under the US-Japan collaboration program, the JAERI Fuel Cleanup System'' (JFCU) was designed and fabricated by JAERI/TPL for testing at the Tritium Systems Test Assembly (TSTA) in Los Alamos National Laboratory as a major subsystem of the simulated fusion fuel cycle. The JFCU was installed in the TSTA in early 1990.

  5. Irradiation Test of Fuel Containing Minor Actinides in the Experimental Fast Reactor Joyo

    NASA Astrophysics Data System (ADS)

    Soga, Tomonori; Sekine, Takashi; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

    The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP accounting for both prompt and delayed heating components, and then adjusted using E/C for 10B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO2-x or AmO2-x in the (U, Pu)O2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel.

  6. Effects of prevaporized fuel on exhaust emissions of an experimental gas turbine combustor

    NASA Technical Reports Server (NTRS)

    Norgren, C. T.; Ingebo, R. D.

    1973-01-01

    Effects of fuel vaporization on the exhaust emission levels of oxides of nitrogen (NOX), carbon monoxide, total hydrocarbons, and smoke number were obtained in an experimental turbojet combustor segment. Two fuel injector types were used in which liquid ASTM A-1 jet fuel and vapor propane fuel were independently controlled to simulate varying degrees of vaporization. Tests were conducted over a range of inlet-air temperatures from 478 to 700 K (860 to 1260 R), pressures from 4 to 20 atmospheres, and combustor reference velocities from 15.3 to 27.4 m/sec (50 to 90 ft/sec). Converting from liquid to complete vapor fuel resulted in NOX reductions as much as 22 percent and smoke number reductions up to 51 percent.

  7. A multi-electrode continuous flow microbial fuel cell with separator electrode assembly design.

    PubMed

    Ahn, Yongtae; Logan, Bruce E

    2012-03-01

    Scaling up microbial fuel cells (MFCs) requires the development of compact reactors with multiple electrodes. A scalable single chamber MFC (130 mL), with multiple graphite fiber brush anodes and a single air-cathode cathode chamber (27 m2/m3), was designed with a separator electrode assembly (SEA) to minimize electrode spacing. The maximum voltage produced in fed-batch operation was 0.65 V (1,000 Ω) with a textile separator, compared to only 0.18 V with a glass fiber separator due to short-circuiting by anode bristles through this separator with the cathode. The maximum power density was 975 mW/m2, with an overall chemical oxygen demand (COD) removal of >90% and a maximum coulombic efficiency (CE) of 53% (50 Ω resistor). When the reactor was switched to continuous flow operation at a hydraulic retention time (HRT) of 8 h, the cell voltage was 0.21 ± 0.04 V, with a very high CE = 85%. Voltage was reduced to 0.13 ± 0.03 V at a longer HRT = 16 h due to a lower average COD concentration, and the CE (80%) decreased slightly with increased oxygen intrusion into the reactor per amount of COD removed. Total internal resistance was 33 Ω, with a solution resistance of 2 Ω. These results show that the SEA type MFC can produce stable power and a high CE, making it useful for future continuous flow treatment using actual wastewaters.

  8. Graphene-Supported Platinum Catalyst-Based Membrane Electrode Assembly for PEM Fuel Cell

    NASA Astrophysics Data System (ADS)

    Devrim, Yilser; Albostan, Ayhan

    2016-08-01

    The aim of this study is the preparation and characterization of a graphene-supported platinum (Pt) catalyst for proton exchange membrane fuel cell (PEMFC) applications. The graphene-supported Pt catalysts were prepared by chemical reduction of graphene and chloroplatinic acid (H2PtCl6) in ethylene glycol. X-ray powder diffraction, thermogravimetric analysis (TGA) and scanning electron microscopy have been used to analyze structure and surface morphology of the graphene-supported catalyst. The TGA results showed that the Pt loading of the graphene-supported catalyst was 31%. The proof of the Pt particles on the support surfaces was also verified by energy-dispersive x-ray spectroscopy analysis. The commercial carbon-supported catalyst and prepared Pt/graphene catalysts were used as both anode and cathode electrodes for PEMFC at ambient pressure and 70°C. The maximum power density was obtained for the Pt/graphene-based membrane electrode assembly (MEA) with H2/O2 reactant gases as 0.925 W cm2. The maximum current density of the Pt/graphene-based MEA can reach 1.267 and 0.43 A/cm2 at 0.6 V with H2/O2 and H2/air, respectively. The MEA prepared by the Pt/graphene catalyst shows good stability in long-term PEMFC durability tests. The PEMFC cell voltage was maintained at 0.6 V without apparent voltage drop when operated at 0.43 A/cm2 constant current density and 70°C for 400 h. As a result, PEMFC performance was found to be superlative for the graphene-supported Pt catalyst compared with the Pt/C commercial catalyst. The results indicate the graphene-supported Pt catalyst could be utilized as the electrocatalyst for PEMFC applications.

  9. Electricity producing property and bacterial community structure in microbial fuel cell equipped with membrane electrode assembly.

    PubMed

    Rubaba, Owen; Araki, Yoko; Yamamoto, Shuji; Suzuki, Kei; Sakamoto, Hisatoshi; Matsuda, Atsunori; Futamata, Hiroyuki

    2013-07-01

    It is important for practical use of microbial fuel cells (MFCs) to not only develop electrodes and proton exchange membranes but also to understand the bacterial community structure related to electricity generation. Four lactate fed MFCs equipped with different membrane electrode assemblies (MEAs) were constructed with paddy field soil as inoculum. The MEAs significantly affected the electricity-generating properties of the MFCs. MEA-I was made with Nafion 117 solution and the other MEAs were made with different configurations of three kinds of polymers. MFC-I equipped with MEA-I exhibited the highest performance with a stable current density of 55 ± 3 mA m⁻². MFC-III equipped with MEA-III with the highest platinum concentration, exhibited the lowest performance with a stable current density of 1.7 ± 0.1 mA m⁻². SEM observation revealed that there were cracks on MEA-III. These results demonstrated that it is significantly important to prevent oxygen-intrusion for improved MFC performance. By comparing the data of DGGE and phylogenetic analyzes, it was suggested that the dominant bacterial communities of MFC-I were constructed with lactate-fermenters and Fe(III)-reducers, which consisted of bacteria affiliated with the genera of Enterobacter, Dechlorosoma, Pelobacter, Desulfovibrio, Propioniferax, Pelosinus, and Firmicutes. A bacterium sharing 100% similarity to one of the DGGE bands was isolated from MFC-I. The 16S rRNA gene sequence of the isolate shared 98% similarity to gram-positive Propioniferax sp. P7 and it was confirmed that the isolate produced electricity in an MFC. These results suggested that these bacteria are valuable for constructing the electron transfer network in MFC.

  10. Synthetic fuel for imitation of municipal solid waste in experimental studies of waste incineration.

    PubMed

    Thipse, S S; Sheng, C; Booty, M R; Magee, R S; Dreizin, E L

    2001-08-01

    Synthetic fuel is prepared to imitate municipal solid waste (MSW) in experimental studies of incineration processes. The fuel is composed based on the Environmental Protection Agency reports on the materials contained in MSW. Uniform synthetic fuel pellets are prepared using available and inexpensive components including newsprint, hardwood mulch, low density polyethylene, iron, animal feed, sand, and water to imitate paperbound, wood, yard trimming, plastic, metal, food wastes, and other materials in MSW. The synthetic fuel preparation procedure enables one to reproduce and modify the fuel for a wide range of experiments in which the mechanisms of waste incineration are addressed. The fuel is characterized using standard ASTM tests and it is shown that its parameters, such as combustion enthalpy, density, as well as moisture, ash and fixed carbon contents are adequate for the representation of municipal solid waste. In addition, chlorine, nitrogen, and sulfur contents of the fuel are shown to be similar to those of MSW. Experiments are conducted in which the synthetic fuel is used for operation of a pilot-scale incinerator research facility. Steady-state temperature operation regimes are achieved and reproduced in these experiments. Thermodynamic equilibrium flame conditions are computed using an isentropic one-dimensional equilibrium code for a wide range of fuel/air ratios. The molecular species used to represent the fuel composition included cellulose, water, iron, polyethylene, methanamine, and silica. The predicted concentrations of carbon monoxide, nitric oxides, and oxygen in the combustion products are compared with the respective experimental concentrations in the pilot-scale incinerator exhaust.

  11. Experimental Study of Unsupported Nonane fuel Droplet Combustion in Microgravity

    NASA Technical Reports Server (NTRS)

    Callahan, B. J.; Avedisian, C. T.; Hertzog, D. E.; Berkery, J. W.

    1999-01-01

    Soot formation in droplet flames is the basic component of the particulate emission process that occurs in spray combustion. The complexity of soot formation motivates a one-dimensional transport condition which has obvious advantages in modeling. Recent models of spherically symmetric droplet combustion have made this assumption when incorporating such aspects as detailed chemistry and radiation. Interestingly, spherical symmetry does not necessarily restrict the results because it has been observed that the properties of carbon formed in flames are not strongly affected by the nature of the fuel or flaming configuration. What is affected, however, are the forces acting on the soot aggregates and where they are trapped by a balance of drag and thermophoretic forces. The distribution of these forces depends on the transport conditions of the flame. Prior studies of spherical droplet flames have examined the droplet burning history of alkanes, alcohols and aromatics. Data are typically the evolution of droplet, flame, extinction, and soot shell diameters. These data are only now just beginning to find their way into comprehensive numerical models of droplet combustion to test proposed oxidation schemes for fuels such as methanol and heptane. In the present study, we report new measurements on the burning history of unsupported nonane droplets in a convection-free environment to promote spherical symmetry. The far-field gas is atmospheric pressure air at room temperature. The evolution of droplet diameter was measured using high speed cine photography of a spark-ignited, droplet within a confined volume in a drop tower. The initial droplet diameters varied between 0.5 mm and 0.6 mm. The challenge of unsupported droplets is to form, deploy and ignite them with minimal disturbance, and then to keep them in the camera field of view. Because of the difficulty of this undertaking, more sophisticated diagnostics for studying soot than photographic were not used. Supporting

  12. CRITICAL CONFIGURATION FOR BERYLLIUM REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH AND 7-TUBE CLUSTERS)

    SciTech Connect

    Margaret A. Marshall

    2012-05-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of 253 unmoderated stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.” The first two experiments in the series were evaluated in HEU-COMP-FAST-001 (SCCA-FUND-EXP-001) and HEU-COMP-FAST-002 (SCCA-FUND-EXP-002). The first experiment had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank (References 1 and 2). The second experiment in the series, performed in early 1963, had the 253 fuel tubes at a 1.506-cm triangular lattice in a 25.96 cm OD core tank and graphite reflectors on all sides. The third set of experiments in the series, performed in mid-1963, which is studied in this evaluation, used beryllium reflectors. The beryllium reflected system was the preferred reactor configuration for this application because of the small thickness of the reflector. The two core configurations had the 253 fuel tubes

  13. Experimental Investigations on Conventional and Semi-Adiabatic Diesel Engine Using Simarouba Biodiesel as Fuel

    NASA Astrophysics Data System (ADS)

    Ravi, M. U.; Reddy, C. P.; Ravindranath, K.

    2013-04-01

    In view of fast depletion of fossil fuels and the rapid rate at which the fuel consumption is taking place all over the world, scientists are searching for alternate fuels for maintaining the growth industrially and economically. Hence search for alternate fuel(s) has become imminent. Out of the limited options for internal combustion engines, the bio diesel fuel appears to be the best. Many advanced countries are implementing several biodiesel initiatives and developmental programmes in order to become self sufficient and reduce the import bills. Biodiesel is biodegradable and renewable fuel with the potential to enhance the performance and reduce engine exhaust emissions. This is due to ready usage of existing diesel engines, fuel distribution pattern, reduced emission profiles, and eco-friendly properties of biodiesel. Simarouba biodiesel (SBD), the methyl ester of Simarouba oil is one such alternative fuel which can be used as substitute to conventional petro-diesel. The present work involves experimental investigation on the use of SBD blends as fuel in conventional diesel engine and semi-adiabatic diesel engine. The oil was triple filtered to eliminate particulate matter and then transesterified to obtain biodiesel. The project envisaged aims at conducting analysis of diesel with SBD blends (10, 20, 30 and 40 %) in conventional engine and semi-adiabatic engine. Also it was decided to vary the injection pressure (180, 190 and 200 bar) and observe its effect on performance and also suggest better value of injection pressure. The engine was made semi adiabatic by coating the piston crown with partially stabilized zirconia (PSZ). Kirloskar AV I make (3.67 kW) vertical, single cylinder, water cooled diesel engine coupled to an eddy current dynamometer with suitable measuring instrumentation/accessories used for the study. Experiments were initially carried out using pure diesel fuel to provide base line data. The test results were compared based on the performance

  14. Design and Development of Membrane Electrode Assembly for Proton Exchange Membrane Fuel Cell

    NASA Astrophysics Data System (ADS)

    Kasat, Harshal Anil

    This work aimed to characterize and optimize the variables that influence the Gas Diffusion Layer (GDL) preparation using design of experiment (DOE) approach. In the process of GDL preparation, the quantity of carbon support and Teflon were found to have significant influence on the Proton Exchange Membrane Fuel Cell (PEMFC). Characterization methods like surface roughness, wetting characteristics, microstructure surface morphology, pore size distribution, thermal conductivity of GDLs were examined using laser interferometer, Goniometer, SEM, porosimetry and thermal conductivity analyzer respectively. The GDLs were evaluated in single cell PEMFC under various operating conditions of temperature and relative humidity (RH) using air as oxidant. Electrodes were prepared with different PUREBLACKRTM and poly-tetrafluoroethylene (PTFE) content in the diffusion layer and maintaining catalytic layer with a Pt-loading (0.4 mg cm-2). In the study, a 73.16 wt.% level of PB and 34 wt.% level of PTFE was the optimal compositions for GDL at 70°C for 70% RH under air atmosphere. For most electrochemical processes the oxygen reduction is very vita reaction. Pt loading in the electrocatalyst contributes towards the total cost of electrochemical devices. Reducing the Pt loading in electrocatalysts with high efficiency is important for the development of fuel cell technologies. To this end, this thesis work reports the approach to lower down the Pt loading in electrocatalyst based on N-doped carbon nanotubes derived from Zeolitic Imidazolate Frameworks (ZIF-67) for oxygen reduction. This electrocatalyst perform with higher electrocatalytic activity and stability for oxygen reduction in fuel cell testing. The electrochemical properties are mainly due to the synergistic effect from N-doped carbon nanotubes derived from ZIF and Pt loading. The strategy with low Pt loading forecasts in emerging highly active and less expensive electrocatalysts in electrochemical energy devices. This

  15. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    NASA Astrophysics Data System (ADS)

    Savander, V. I.; Shumskiy, B. E.; Pinegin, A. A.

    2016-12-01

    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

  16. Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II (Experimental Breeder Reactor)

    SciTech Connect

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1988-01-01

    The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs.

  17. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    NASA Astrophysics Data System (ADS)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  18. Dimensionless numbers and correlating equations for the analysis of the membrane-gas diffusion electrode assembly in polymer electrolyte fuel cells

    NASA Astrophysics Data System (ADS)

    Gyenge, E. L.

    The Quraishi-Fahidy method [Can. J. Chem. Eng. 59 (1981) 563] was employed to derive characteristic dimensionless numbers for the membrane-electrolyte, cathode catalyst layer and gas diffuser, respectively, based on the model presented by Bernardi and Verbrugge for polymer electrolyte fuel cells [AIChE J. 37 (1991) 1151]. Monomial correlations among dimensionless numbers were developed and tested against experimental and mathematical modeling results. Dimensionless numbers comparing the bulk and surface-convective ionic conductivities, the electric and viscous forces and the current density and the fixed surface charges, were employed to describe the membrane ohmic drop and its non-linear dependence on current density due to membrane dehydration. The analysis of the catalyst layer yielded electrode kinetic equivalents of the second Damköhler number and Thiele modulus, influencing the penetration depth of the oxygen reduction front based on the pseudohomogeneous film model. The correlating equations for the catalyst layer could describe in a general analytical form, all the possible electrode polarization scenarios such as electrode kinetic control coupled or not with ionic and/or oxygen mass transport limitation. For the gas diffusion-backing layer correlations are presented in terms of the Nusselt number for mass transfer in electrochemical systems. The dimensionless number-based correlating equations for the membrane electrode assembly (MEA) could provide a practical approach to quantify single-cell polarization results obtained under a variety of experimental conditions and to implement them in models of the fuel cell stack.

  19. Portable instrument for inspecting irradiated nuclear-fuel assemblies in a water-filled storage pond by measurement of induced Cerenkov radiation

    DOEpatents

    Nicholson, N.; Dowdy, E.J.; Holt, D.M.; Stump, C.J. Jr.

    1982-05-13

    A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

  20. Critical Configuration and Physics Mesaurements for Graphite Reflected Assemblies of U(93.15)O2 Fuel Rods (1.27-CM Pitch)

    SciTech Connect

    Margaret A. Marshall

    2011-09-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory's Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950's efforts were made to study 'power plants for the production of electrical power in space vehicles'. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in FY 1964, 1965, and 1966. A summary of the program's effort was compiled in 1967. The delayed critical experiments served as a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated 253 stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. 'The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.' The experiment studied within this evaluation was the first of the series and had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Information for this evaluation was compiled from Reference 1 and 2, reports on subsequent experiments in the series, and the experimental logbook as well as from communication with the experimenter, John T. Mihalczo.

  1. Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan

    SciTech Connect

    Adkins, Harold E.

    2013-04-01

    Under current U.S. Nuclear Regulatory Commission regulation, it is not sufficient for used nuclear fuel (UNF) to simply maintain its integrity during the storage period, it must maintain its integrity in such a way that it can withstand the physical forces of handling and transportation associated with restaging the fuel and moving it to treatment or recycling facilities, or a geologic repository. Hence it is necessary to understand the performance characteristics of aged UNF cladding and ancillary components under loadings stemming from transport initiatives. Researchers would like to demonstrate that enough information, including experimental support and modeling and simulation capabilities, exists to establish a preliminary determination of UNF structural performance under normal conditions of transport (NCT). This research, development and demonstration (RD&D) plan describes a methodology, including development and use of analytical models, to evaluate loading and associated mechanical responses of UNF rods and key structural components. This methodology will be used to provide a preliminary assessment of the performance characteristics of UNF cladding and ancillary components under rail-related NCT loading. The methodology couples modeling and simulation and experimental efforts currently under way within the Used Fuel Disposition Campaign (UFDC). The methodology will involve limited uncertainty quantification in the form of sensitivity evaluations focused around available fuel and ancillary fuel structure properties exclusively. The work includes collecting information via literature review, soliciting input/guidance from subject matter experts, performing computational analyses, planning experimental measurement and possible execution (depending on timing), and preparing a variety of supporting documents that will feed into and provide the basis for future initiatives. The methodology demonstration will focus on structural performance evaluation of

  2. Experimental investigation and modeling of an aircraft Otto engine operating with gasoline and heavier fuels

    NASA Astrophysics Data System (ADS)

    Saldivar Olague, Jose

    A Continental "O-200" aircraft Otto-cycle engine has been modified to burn diesel fuel. Algebraic models of the different processes of the cycle were developed from basic principles applied to a real engine, and utilized in an algorithm for the simulation of engine performance. The simulation provides a means to investigate the performance of the modified version of the Continental engine for a wide range of operating parameters. The main goals of this study are to increase the range of a particular aircraft by reducing the specific fuel consumption of the engine, and to show that such an engine can burn heavier fuels (such as diesel, kerosene, and jet fuel) instead of gasoline. Such heavier fuels are much less flammable during handling operations making them safer than aviation gasoline and very attractive for use in flight operations from naval vessels. The cycle uses an electric spark to ignite the heavier fuel at low to moderate compression ratios, The stratified charge combustion process is utilized in a pre-chamber where the spray injection of the fuel occurs at a moderate pressure of 1200 psi (8.3 MPa). One advantage of fuel injection into the combustion chamber instead of into the intake port, is that the air-to-fuel ratio can be widely varied---in contrast to the narrower limits of the premixed combustion case used in gasoline engines---in order to obtain very lean combustion. Another benefit is that higher compression ratios can be attained in the modified cycle with heavier fuels. The combination of injection into the chamber for lean combustion, and higher compression ratios allow to limit the peak pressure in the cylinder, and to avoid engine damage. Such high-compression ratios are characteristic of Diesel engines and lead to increase in thermal efficiency without pre-ignition problems. In this experimental investigation, operations with diesel fuel have shown that considerable improvements in the fuel efficiency are possible. The results of

  3. High Performance Fuel Cell and Electrolyzer Membrane Electrode Assemblies (MEAs) for Space Energy Storage Systems

    NASA Technical Reports Server (NTRS)

    Valdez, Thomas I.; Billings, Keith J.; Kisor, Adam; Bennett, William R.; Jakupca, Ian J.; Burke, Kenneth; Hoberecht, Mark A.

    2012-01-01

    Regenerative fuel cells provide a pathway to energy storage system development that are game changers for NASA missions. The fuel cell/ electrolysis MEA performance requirements 0.92 V/ 1.44 V at 200 mA/cm2 can be met. Fuel Cell MEAs have been incorporated into advanced NFT stacks. Electrolyzer stack development in progress. Fuel Cell MEA performance is a strong function of membrane selection, membrane selection will be driven by durability requirements. Electrolyzer MEA performance is catalysts driven, catalyst selection will be driven by durability requirements. Round Trip Efficiency, based on a cell performance, is approximately 65%.

  4. Numerical Investigation on Sensitivity of Liquid Jet Breakup to Physical Fuel Properties with Experimental Comparison

    NASA Astrophysics Data System (ADS)

    Kim, Dokyun; Bravo, Luis; Matusik, Katarzyna; Duke, Daniel; Kastengren, Alan; Swantek, Andy; Powell, Christopher; Ham, Frank

    2016-11-01

    One of the major concerns in modern direct injection engines is the sensitivity of engine performance to fuel characteristics. Recent works have shown that even slight differences in fuel properties can cause significant changes in efficiency and emission of an engine. Since the combustion process is very sensitive to the fuel/air mixture formation resulting from disintegration of liquid jet, the precise assessment of fuel sensitivity on liquid jet atomization process is required first to study the impact of different fuels on the combustion. In the present study, the breaking process of a liquid jet from a diesel injector injecting into a quiescent gas chamber is investigated numerically and experimentally for different liquid fuels (n-dodecane, iso-octane, CAT A2 and C3). The unsplit geometric Volume-of-Fluid method is employed to capture the phase interface in Large-eddy simulations and results are compared against the radiography measurement from Argonne National Lab including jet penetration, liquid mass distribution and volume fraction. The breakup characteristics will be shown for different fuels as well as droplet PDF statistics to demonstrate the influences of the physical properties on the primary atomization of liquid jet. Supported by HPCMP FRONTIER award, US DOD, Office of the Army.

  5. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect

    Ilas, Germina; Gauld, Ian C

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  6. Summary of experimental data for critical arrays of water moderated Fast Test Reactor fuel

    SciTech Connect

    Durst, B.M.; Bierman, S.R.; Clayton, E.D.; Mincey, J.F.; Primm, R.T. III

    1981-05-01

    A research program, funded by the Consolidated Fuel Reprocessing Program (CFRP) of Oak Ridge National Laboratory (ORNL), was initiated at Battelle Pacific Northwest Laboratory (PNL) to acquire experimental data on heterogeneous water moderated arrays of Fast Test Reactor (FTR) fuel pins. The objective of this program is to provide critical experiment data for validating calculational techniques used in criticality assessments of reprocessing equipment containing FTR-type fuels. Consequently, the experiments were designed to permit accurate definition in Monte Carlo computer codes currently used in these assessments. Square and triangular pitched lattices of fuel have been constructed under a variety of conditions covering the range from undermoderated to overmoderated arrays. Experiments were conducted composed of arrays which were water reflected, partially concrete reflected, and arrays with interspersed solid neutron absorbers. The absorbers utilized were Boral, and cadmium plates and gadolinium cylindrical rods. Data from non-CFRP sponsored subcritical experiments (previously performed at Hanford) also are included.

  7. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study

    SciTech Connect

    Ham, Y S; Sitaraman, S

    2008-12-24

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided

  8. A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Bushuev, A. V.; Kozhin, A. F.; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E.

    2016-12-01

    An active neutron method for measuring the residual mass of 235U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual 235U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of 238U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

  9. Fabrication of gas impervious edge seal for a bipolar gas distribution assembly for use in a fuel cell

    DOEpatents

    Kaufman, Arthur; Werth, John

    1986-01-01

    A bipolar gas reactant distribution assembly for use in a fuel cell is disclosed, the assembly having a solid edge seal to prevent leakage of gaseous reactants wherein a pair of porous plates are provided with peripheral slits generally parallel to, and spaced apart from two edges of the plate, the slit being filled with a solid, fusible, gas impervious edge sealing compound. The plates are assembled with opposite faces adjacent one another with a layer of a fusible sealant material therebetween the slits in the individual plates being approximately perpendicular to one another. The plates are bonded to each other by the simultaneous application of heat and pressure to cause a redistribution of the sealant into the pores of the adjacent plate surfaces and to cause the edge sealing compound to flow and impregnate the region of the plates adjacent the slits and comingle with the sealant layer material to form a continuous layer of sealant along the edges of the assembled plates.

  10. An experimental and modeling study of fires in ventilated ducts; Part 1: Liquid fuels

    SciTech Connect

    Comitis, S.C.; Glasser, D.; Young, B.D. . Dept. of Chemical Engineering)

    1994-03-01

    A theoretical model for fire propagation through a fuel-lined duct with a radially well-mixed axial flow is presented. The gas-phase is modeled as a steady-state process whereas the condensed-phase (fuel source) is taken to be the cause of transient fire propagation along the duct. Experiments were performed in a small-scale duct where fire propagation and gas temperature histories were acquired. Experimental results confirm hypotheses of pseudo-steady-state gas-phase processes. Theory and experiment display transient fire propagation for typical duct fire scenarios where initial fuel mass loading is constant with respect to duct length. The phenomena observed, as predicted by theory, is an initial jump'' of the fully developed combustion process followed by convergence to a steady-state constant fire propagation speed. The theory is in all important aspects able to quantitatively model the experimental results.

  11. CRITICAL CONFIGURATION AND PHYSICS MEASUREMENTS FOR GRAPHITE REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH)

    SciTech Connect

    Margaret A. Marshall

    2012-03-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of 253 unmoderated stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.” The first experiment in the series was evaluated in HEU-COMP-FAST-001. It had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank (References 1 and 2). The second experiment in the series, performed in early 1963, which is studied in this evaluation, had the 253 fuel tubes at a 1.506-cm triangular lattice in a 25.96 cm OD core tank and graphite reflectors on all sides. The experiment has been determined to represent an acceptable benchmark experiment. Information for this evaluation was compiled from published reports on all three parts of the experimental series (Reference 1-5) and the experimental logbook as

  12. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.

  13. Assembly of a Cost-Effective Anode Using Palladium Nanoparticles for Alkaline Fuel Cell Applications

    ERIC Educational Resources Information Center

    Feliciano-Ramos, Ileana; Casan~as-Montes, Barbara; García-Maldonado, María M.; Menendez, Christian L.; Mayol, Ana R.; Díaz-Vazquez, Liz M.; Cabrera, Carlos R.

    2015-01-01

    Nanotechnology allows the synthesis of nanoscale catalysts, which offer an efficient alternative for fuel cell applications. In this laboratory experiment, the student selects a cost-effective anode for fuel cells by comparing three different working electrodes. These are commercially available palladium (Pd) and glassy carbon (GC) electrodes, and…

  14. Long-term changes in community assembly, resistance, and resilience following experimental floods.

    PubMed

    Robinson, Christopher T

    2012-10-01

    This study examined the long-term changes in community assembly, resistance, and resilience of macroinvertebrates following 10 years of experimental floods in a flow regulated river. Physico-chemistry, macroinvertebrates, and periphyton biomass were monitored before and sequentially after each of 22 floods, and drift/seston was collected during six separate floods over the study period. The floods reduced the density and taxon richness of macroinvertebrates, and a nonmetric dimensional scaling (NMDS) analysis distinguished temporal shifts in community assembly. Resistance (measured as the relative lack of loss in density) tofloods varied among taxa, and the abundance of resistant taxa was related to the temporal changes in community assembly. Community resistance was inversely related to flood magnitude with all larger floods (> 25 m3/s, > 16-fold over baseflow) reducing densities by > 75% regardless of flood year, whereas smaller floods (< 20 m3/s) reduced taxon richness approximately twofold less than larger floods. No relationship was found between flood magnitude and the relative loss in periphyton biomass. Resilience was defined as the recovery slope (positive slope of a parameter with time following each flood) and was unrelated to shifts in community assembly or resistance. Macroinvertebrate drift and seston demonstrated hysteresis (i.e., a temporal response in parameter quantity with change in discharge) during each flood, although larger floods typically had two peaks in both parameters. The first peak was a response to the initial increases in flow, whereas the second peak was associated with streambed disturbance (substrate mobility) and side-slope failure causing increased scour. Drift density was 3-9 times greater and that of seston 3-30 times greater during larger floods than smaller floods. These results demonstrate temporal shifts in macroinvertebrate community assembly toward a pre-dam assemblage following sequential floods in this flow regulated

  15. Experimental Measurement and Numerical Modeling of the Effective Thermal Conductivity of TRISO Fuel Compacts

    SciTech Connect

    Folsom, Charles; Xing, Changhu; Jensen, Colby; Ban, Heng; Marshall, Douglas W.

    2015-03-01

    Accurate modeling capability of thermal conductivity of tristructural-isotropic (TRISO) fuel compacts is important to fuel performance modeling and safety of Generation IV reactors. To date, the effective thermal conductivity (ETC) of tristructural-isotropic (TRISO) fuel compacts has not been measured directly. The composite fuel is a complicated structure comprised of layered particles in a graphite matrix. In this work, finite element modeling is used to validate an analytic ETC model for application to the composite fuel material for particle-volume fractions up to 40%. The effect of each individual layer of a TRISO particle is analyzed showing that the overall ETC of the compact is most sensitive to the outer layer constituent. In conjunction with the modeling results, the thermal conductivity of matrix-graphite compacts and the ETC of surrogate TRISO fuel compacts have been successfully measured using a previously developed measurement system. The ETC of the surrogate fuel compacts varies between 50 and 30 W m-1 K-1 over a temperature range of 50-600°C. As a result of the numerical modeling and experimental measurements of the fuel compacts, a new model and approach for analyzing the effect of compact constituent materials on ETC is proposed that can estimate the fuel compact ETC with approximately 15-20% more accuracy than the old method. Using the ETC model with measured thermal conductivity of the graphite matrix-only material indicate that, in the composite form, the matrix material has a much greater thermal conductivity, which is attributed to the high anisotropy of graphite thermal conductivity. Therefore, simpler measurements of individual TRISO compact constituents combined with an analytic ETC model, will not provide accurate predictions of overall ETC of the compacts emphasizing the need for measurements of composite, surrogate compacts.

  16. Experimental test plan: USDOE/JAERI collaborative program for the coated particle fuel performance test

    SciTech Connect

    Kania, M.J.; Fukuda, K.

    1989-12-01

    This document describes the coated-particle fuel performance test agreed to under Annex 2 of the arrangement between the US Department of Energy and the Japan Atomic Energy Research Institute on cooperation in research and development regarding high-temperature gas-cooled reactors (HTGRs). The test will evaluate the behavior of reference fuel compacts containing coated-particle fuels fabricated according to the specifications for the US Modular HTGR and the Japanese High-Temperature Engineering Test Reactor (HTTR) concepts. Two experimental capsules, HRB-21 and HRB-22, are being tested. Capsule HRB-21 contains only US reference fuel, and HRB-22 contains only JAERI reference fuel. Both capsules will be irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Capsule HRB-21 will be operated at a mean volumetric fuel temperature of 975{degrees}C and will achieve a peak fissile burnup of 26% fissions per initial metal atom (FIMA) and a fast fluence of {le}4.5 {times} 10{sup 25} neutrons/m{sup 2}. Capsule HRB-22 will be operated at a mean centerline fuel temperature of 1250 to 1300{degrees}C and will achieve a peak fissile burnup of 5.5% FIMA and a fast fluence of 1.7 {times} 10{sup 25} neutrons/m{sup 2}. Performance of the fuels during irradiation will be closely monitored using on-line fission gas surveillance. Following irradiation, both capsules will undergo detailed examinations and core heatup simulation testing. Results from in-reactor monitoring and postirradiation testing will be analyzed to comparatively assess US and Japanese coated-particle fuel performance. 3 refs., 9 figs., 10 tabs.

  17. An Experimental Investigation of Hypergolic Ignition Delay of Hydrogen Peroxide with Fuel Mixtures

    NASA Technical Reports Server (NTRS)

    Blevins, John A.; Gostowski, Rudy; Chianese, Silvio

    2003-01-01

    An experimental evaluation of decomposition and ignition delay of hydrogen peroxide at concentrations of 80% to 98% with combinations of hydrocarbon fuels, tertiary amines and transition metal chelates will be presented in the proposed paper. The results will be compared to hydrazine ignition delays with hydrogen peroxide and nitric acid mixtures using the same test apparatus.

  18. Experimental hydrogen-fueled automotive engine design data-base project. Volume 2. Main technical report

    SciTech Connect

    Swain, M.R.; Adt, R.R. Jr.; Pappas, J.M.

    1983-05-01

    Operational performance and emissions characteristics of hydrogen-fueled engines are reviewed. The project activities are reviewed including descriptions of the test engine and its components, the test apparatus, experimental techniques, experiments performed and the results obtained. Analyses of other hydrogen engine project data are also presented and compared with the results of the present effort.

  19. Instrumented fuels test for FFTF

    SciTech Connect

    Feigenbutz, L.V.; Hoth, C.W.

    1980-01-01

    In support of the LMFBR Fuels Development Program, Hanford Engineering Development Laboratory (HEDL) has designed the Fuels Open Test Assembly (FOTA) for fuels testing at the Fast Flux Test Facility (FFTF). The FOTA is a test vehicle designed to contain and support instrumented fuel experiments in the Fast Test Reactor (FTR) at FFTF. The initial two FOTA experiments will characterize the reference Driver Fuel Assembly performance in the FTR and provide experimental data to evaluate thermohydraulic models used to predict assembly performance. The design features and fabrication are described for the first two FOTA instrumented fuel experiments, which have been fabricated and are now in the FTR. A brief description of the FOTA test vehicle is included.

  20. Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

    NASA Astrophysics Data System (ADS)

    Lubina, A. S.; Subbotin, A. S.; Sedov, A. A.; Frolov, A. A.

    2016-12-01

    The fast sodium reactor fuel assembly (FA) with U-Pu-Zr metallic fuel is described. In comparison with a "classical" fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

  1. High-areal-density fuel assembly in direct-drive cryogenic implosions.

    PubMed

    Sangster, T C; Goncharov, V N; Radha, P B; Smalyuk, V A; Betti, R; Craxton, R S; Delettrez, J A; Edgell, D H; Glebov, V Yu; Harding, D R; Jacobs-Perkins, D; Knauer, J P; Marshall, F J; McCrory, R L; McKenty, P W; Meyerhofer, D D; Regan, S P; Seka, W; Short, R W; Skupsky, S; Soures, J M; Stoeckl, C; Yaakobi, B; Shvarts, D; Frenje, J A; Li, C K; Petrasso, R D; Séguin, F H

    2008-05-09

    The first observation of ignition-relevant areal-density deuterium from implosions of capsules with cryogenic fuel layers at ignition-relevant adiabats is reported. The experiments were performed on the 60-beam, 30-kJUV OMEGA Laser System [T. R. Boehly, Opt. Commun. 133, 495 (1997)10.1016/S0030-4018(96)00325-2]. Neutron-averaged areal densities of 202+/-7 mg/cm2 and 182+/-7 mg/cm2 (corresponding to estimated peak fuel densities in excess of 100 g/cm3) were inferred using an 18-kJ direct-drive pulse designed to put the converging fuel on an adiabat of 2.5. These areal densities are in good agreement with the predictions of hydrodynamic simulations indicating that the fuel adiabat can be accurately controlled under ignition-relevant conditions.

  2. Assemblies with both target and fuel pins in an isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

  3. Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies

    NASA Astrophysics Data System (ADS)

    Gore, B. F.; McNair, G. W.; Heaberlin, S. W.

    1980-05-01

    Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuels disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close packing cannot be achieved due to fuel rod bowing. It is concluded the disposal canisters should be sized on the basis of assumed optimum moderation.

  4. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    SciTech Connect

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the STVU content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The STYU content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared.

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

    NASA Technical Reports Server (NTRS)

    Vezhba, I.

    1979-01-01

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

  6. Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations; Yucca Mountain Site Characterization Project

    SciTech Connect

    Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J.

    1991-12-01

    This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 {times} 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990.

  7. Trapped Vortex Combustion Chamber: Design and Experimental Investigations Using Hydrogen as Fuel

    NASA Astrophysics Data System (ADS)

    Kulshreshtha, D. B.; Channiwala, S. A.

    2014-01-01

    The design of trapped vortex combustion chamber was undertaken as a part of ongoing research on micro combustion chamber using hydrogen as fuel. The reacting experimental studies were then carried out on the designed chamber. The fuel was injected directly into the cavity. The combustion was first initiated in the cavity with 3 % of the main flow air supplied in reverse direction to the fuel flow. The combustion in cavity was of rich type. Temperature levels in the range of 900 K were encountered in the cavity. Thereafter, diffusion combustion was initiated using the flame generated in the cavity. The temperature levels in this stage were in the range of 1,800 K. The overall pressure drop for a trapped vortex combustor was less than 5 % at all operating parameters.

  8. Experimental and theoretical investigations in stimuli responsive dendrimer-based assemblies

    NASA Astrophysics Data System (ADS)

    Molla, Mijanur Rahaman; Rangadurai, Poornima; Pavan, Giovanni M.; Thayumanavan, S.

    2015-02-01

    Stimuli-responsive macromolecular assemblies are of great interest in drug delivery applications, as it holds the promise to keep the drug molecules sequestered under one set of conditions and release them under another. The former set of conditions could represent circulation, while the latter could represent a disease location. Over the past two decades, sizeable contributions to this field have come from dendrimers, which along with their monodispersity, provide great scope for structural modifications at the molecular level. In this paper, we briefly discuss the various synthetic strategies that have been developed so far to obtain a range of functional dendrimers. We then discuss the design strategies utilized to introduce stimuli responsive elements within the dendritic architecture. The stimuli itself are broadly classified into two categories, viz. extrinsic and intrinsic. Extrinsic stimuli are externally induced such as temperature and light variations, while intrinsic stimuli involve physiological aberrations such as variations in pH, redox conditions, proteins and enzyme concentrations in pathological tissues. Furthermore, the unique support from molecular dynamics (MD) simulations has been highlighted. MD simulations have helped back many of the observations made from assembly formation properties to rationalized the mechanism of drug release and this has been illustrated with discussions on G4 PPI (Poly propylene imine) dendrimers and biaryl facially amphiphilic dendrimers. The synergy that exists between experimental and theoretical studies open new avenues for the use of dendrimers as versatile drug delivery systems.

  9. Experimental and Theoretical Investigations in Stimuli Responsive Dendrimer-based Assemblies

    PubMed Central

    Molla, Mijanur Rahaman; Rangadurai, Poornima

    2014-01-01

    Stimuli-responsive macromolecular assemblies are of great interest in drug delivery applications, as it holds the promise to keep the drug molecules sequestered under one set of conditions and release them under another. The former set of conditions could represent circulation, while the latter could represent a disease location. Over the past two decades, sizeable contributions to this field have come from dendrimers, which along with their monodispersity, provide great scope for structural modifications at the molecular level. In this paper, we briefly discuss the various synthetic strategies that have been developed so far to obtain a range of functional dendrimers. We then discuss the design strategies utilized to introduce stimuli responsive elements within the dendritic architecture. The stimuli itself are broadly classified into two categories, viz. extrinsic and intrinsic. Extrinsic stimuli are externally induced such as temperature and light variations, while intrinsic stimuli involve physiological aberrations such as variations in pH, redox conditions, proteins and enzyme concentrations in pathological tissues. Furthermore, the unique support from molecular dynamics (MD) simulations has been highlighted. MD simulations have helped back many of the observations made from assembly formation properties to rationalized the mechanism of drug release and this has been illustrated with discussions on G4 PPI (Poly propylene imine) dendrimers and biaryl facially amphiphilic dendrimers. The synergy that exists between experimental and theoretical studies open new avenues for the use of dendrimers as versatile drug delivery systems. PMID:25260107

  10. Templated assembly of photoswitches significantly increases the energy-storage capacity of solar thermal fuels.

    PubMed

    Kucharski, Timothy J; Ferralis, Nicola; Kolpak, Alexie M; Zheng, Jennie O; Nocera, Daniel G; Grossman, Jeffrey C

    2014-05-01

    Large-scale utilization of solar-energy resources will require considerable advances in energy-storage technologies to meet ever-increasing global energy demands. Other than liquid fuels, existing energy-storage materials do not provide the requisite combination of high energy density, high stability, easy handling, transportability and low cost. New hybrid solar thermal fuels, composed of photoswitchable molecules on rigid, low-mass nanostructures, transcend the physical limitations of molecular solar thermal fuels by introducing local sterically constrained environments in which interactions between chromophores can be tuned. We demonstrate this principle of a hybrid solar thermal fuel using azobenzene-functionalized carbon nanotubes. We show that, on composite bundling, the amount of energy stored per azobenzene more than doubles from 58 to 120 kJ mol(-1), and the material also maintains robust cyclability and stability. Our results demonstrate that solar thermal fuels composed of molecule-nanostructure hybrids can exhibit significantly enhanced energy-storage capabilities through the generation of template-enforced steric strain.

  11. Experimental research on water management in proton exchange membrane fuel cells

    NASA Astrophysics Data System (ADS)

    Yu, Li-jun; Chen, Wen-can; Qin, Ming-jun; Ren, Geng-po

    A simulated cathode flow channel experiment system was set up based on the gas flow rate and water flow rate in the PEM fuel cell. With the assistance of the visualization system, high-sensitivity double parallel conductance probes flow regime inspecting technique was adopted successfully in the experiment system to inspect the flow regime of the gas-liquid two-phase flow in the PEM fuel cell. The research results show that the double parallel conductance probes inspecting system and the flow regime image system for the gas-liquid two-phase flow in the PEM fuel cell simulated channel both can judge the slug flow and annular flow in it, and the double parallel conductance probes flow regime inspecting system can divide the annular flow into three subtypes. The main probes inspecting system and the assistant image system validate reciprocally, which enhances the experimental veracity. The typical flow regimes of the PEM fuel cell simulated channel include slug flow, annular flow with big water film wave, annular flow with small water film wave and annular flow without water film wave. With the increase of the liquid superficial velocity, the frequencies of liquid slug and wave of liquid film increase. The flow regime map in the flow channel of the PEM fuel cell was developed. The flow regime of the gas-liquid two-phase flow in a PEM fuel cell in different operating conditions can be forecasted with this map. With the PEM fuel cell operating condition in this study, the flow regimes of gas-liquid two-phase flow for different cases are all annular flow with small water film wave, and the liquid film waves more with bigger current density. With the location closer to the channel outlet, the liquid film waves are more for the same current density.

  12. Method for removing solid particulate material from within liquid fuel injector assemblies

    DOEpatents

    Simandl, Ronald F.; Brown, John D.; Andriulli, John B.; Strain, Paul D.

    1998-01-01

    A method for removing residual solid particulate material from the interior of liquid fuel injectors and other fluid flow control mechanisms having or being operatively associated with a flow-regulating fixed or variable orifice. The method comprises the sequential and alternate introduction of columns of a non-compressible liquid phase and columns of a compressed gas phase into the body of a fuel injector whereby the expansion of each column of the gas phase across the orifice accelerates the liquid phase in each trailing column of the liquid phase and thereby generates turbulence in each liquid phase for lifting and entraining the solid particulates for the subsequent removal thereof from the body of the fuel injector.

  13. Method for removing solid particulate material from within liquid fuel injector assemblies

    DOEpatents

    Simandl, R.F.; Brown, J.D.; Andriulli, J.B.; Strain, P.D.

    1998-09-08

    A method is described for removing residual solid particulate material from the interior of liquid fuel injectors and other fluid flow control mechanisms having or being operatively associated with a flow-regulating fixed or variable orifice. The method comprises the sequential and alternate introduction of columns of a non-compressible liquid phase and columns of a compressed gas phase into the body of a fuel injector whereby the expansion of each column of the gas phase across the orifice accelerates the liquid phase in each trailing column of the liquid phase and thereby generates turbulence in each liquid phase for lifting and entraining the solid particulates for the subsequent removal thereof from the body of the fuel injector. 1 fig.

  14. Experimental analysis, modeling, and optimal control of PEM fuel cell electrochemistry

    NASA Astrophysics Data System (ADS)

    Dhanda, Abhishek

    Polymer Electrolyte Membrane (PEM) fuel cells are touted to play a major role in the green hydrogen based economy. However performance issues need to be addressed for mass commercialization of fuel cells. Besides other factors, slow chemical and electrochemical surface reactions on Pt based catalysts cause large potential loss, and are the primary cause of performance degradation in PEM fuel cells. The kinetics of the oxygen reduction reaction (ORR) at the cathode and the hydrogen oxidation reaction (HOR) at the anode depend on the crystal orientation. Accurate modeling of PEM electrochemistry at the Pt/electrolyte interface requires study of reaction mechanisms on well defined Pt surfaces. In this thesis, electrochemistry on single crystal Pt/Nafion interfaces was studied using a novel experimental setup. Steady state and transient impedance spectroscopy experiments were performed at different operating temperatures. These results are used to derive a kinetic model of the adsorbed species and the overall reaction. Based on such a kinetic model of electrochemical reactions, an approach is presented to improve the time-average performance of PEM fuel cells. Electrochemical kinetic rates depend on operating voltage and current signals. Optimal time varying profile of operating current were derived using variational calculus. Simulation results are presented for demonstrating the application of optimal control approach in reducing carbon monoxide (CO) poisoning in PEM fuel cells.

  15. Experimental studies of thermal and chemical interactions between oxide and silicide nuclear fuels with water

    SciTech Connect

    farahani, A.A.; Corradini, M.L.

    1995-09-01

    Given some transient power/cooling mismatch is a nuclear reactor and its inability to establish the necessary core cooling, energetic fuel-coolant interactions (FCI`s commonly called `vapor explosions`) could occur as a result of the core melting and coolant contact. Although a large number of studies have been done on energetic FCI`s, very few experiments have been performed with the actual fuel materials postulated to be produced in severe accidents. Because of the scarcity of well-characterized FCI data for uranium allows in noncommercial reactors (cermet and silicide fuels), we have conducted a series of experiments to provide a data base for the foregoing materials. An existing 1-D shock-tube facility was modified to handle depleted radioactive materials (U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al). Our objectives have been to determine the effects of the initial fuel composition and temperature and the driving pressure (triggering) on the explosion work output, dynamic pressures, transient temperatures, and the hydrogen production. Experimental results indicate limited energetics, mainly thermal interactions, for these fuel materials as compared to aluminum where more chemical reactions occur between the molten aluminum and water.

  16. Cargo Fire Hazards and Hazard Control for the Supplement Fuel Supply Assembly (SFSA).

    DTIC Science & Technology

    1980-08-01

    spilled as a result of the fuel hose failure. Fire water from the supply ship can be used to disperse the spilled fuel so that a boat can be dispatched...aided by agitating the spill. This agitation can be accomplished with the use of fire water monitor nozzles on a utili- ty/fire boat to be discussed...in the next section. Fire water nozzles available for the craft used to fill the Dracones can also be used for spill control purposes should a spill

  17. Tritium experiments on components for fusion fuel processing at the Tritium Systems Test Assembly

    SciTech Connect

    Konishi, S.; Yoshida, H.; Naruse, Y. ); Carlson, R.V.; Binning, K.E.; Bartlit, J.R.; Anderson, J.L. )

    1990-01-01

    Under a collaborative agreement between US and Japan, two tritium processing components, a palladium diffuser and a ceramic electrolysis cell have been tested with tritium for application to a Fuel Cleanup System (FCU) for plasma exhaust processing at the Los Alamos National Laboratory. The fundamental characteristics, compatibility with tritium, impurities effects with tritium, and long-term behavior of the components, were studied over a three year period. Based on these studies, an integrated process loop, JAERI Fuel Cleanup System'' equipped with above components was installed at the TSTA for full scale demonstration of the plasma exhaust reprocessing.

  18. High Fuel Utilization in Solid Oxide Fuel Cells: Experimental Characterization and Data Analysis with Continuous Wavelet Transform

    NASA Astrophysics Data System (ADS)

    Esposito, Angelo; Russo, Luigi; Kändler, Christoph; Pianese, Cesare; Ludwig, Bastian; Steiner, Nadia Yousfi

    2016-06-01

    The on-line diagnostics of Solid Oxide Fuel Cells (SOFCs) is a critical tool to achieve optimal performance and extend the lifetime. The Continuous Wavelet Transform (CWT) methodology was applied to the SOFC voltage signal to detect signatures that reveal the presence of a fault in the cell/stack. The selected fault was anode re-oxidation caused by high Fuel Utilization (FU) (higher then nominal). To experimentally emulate the high FU faults, a standard test procedure was developed, which was used to characterize a μ-CHP system at high FU operation. To complete the analysis, data collected on Single Cells were exploited too. The CWT was applied to the voltage signal for each FU level to verify the qualitative difference (signature) between the signals at different FU's within the same tests as well as the correspondence between the same conditions over different tests. A statistical study was performed to quantify the observed differences and to determine the correspondence between CWT coefficients and operating conditions. The approach proves to be suitable to diagnose high FU in SOFC, showing a successful detection rate above 76%. The results show the good potential of using the CWT methodology as diagnostic tools for SOFCs from cell to stack level.

  19. Experimental investigation of burnup credit for safe transport, storage, and disposal of spent nuclear fuel.

    SciTech Connect

    Berry, Donald T.; Harms, Gary A.; Ford, John T.; Walker, Sharon Ann; Helmick, Paul H.; Pickard, Paul S.

    2004-04-01

    This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

  20. Experimental Investigation of Burnup Credit for Safe Transport, Storage, and Disposal of Spent Nuclear Fuel

    SciTech Connect

    Harms, Gary A.; Helmick, Paul H.; Ford, John T.; Walker, Sharon A.; Berry, Donald T.; Pickard, Paul S.

    2004-04-01

    This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

  1. Investigation of Ruthenium Dissolution in Advanced Membrane Electrode Assemblies for Direct Methanol Based Fuel Cells Stacks

    NASA Technical Reports Server (NTRS)

    Valdez, T. I.; Firdosy, S.; Koel, B. E.; Narayanan, S. R.

    2005-01-01

    This viewgraph presentation gives a detailed review of the Direct Methanol Based Fuel Cell (DMFC) stack and investigates the Ruthenium that was found at the exit of the stack. The topics include: 1) Motivation; 2) Pathways for Cell Degradation; 3) Cell Duration Testing; 4) Duration Testing, MEA Analysis; and 5) Stack Degradation Analysis.

  2. Experimental investigation on the characteristics of supersonic fuel spray and configurations of induced shock waves

    PubMed Central

    Wang, Yong; Yu, Yu-song; Li, Guo-xiu; Jia, Tao-ming

    2017-01-01

    The macro characteristics and configurations of induced shock waves of the supersonic sprays are investigated by experimental methods. Visualization study of spray shape is carried out with the high-speed camera. The macro characteristics including spray tip penetration, velocity of spray tip and spray angle are analyzed. The configurations of shock waves are investigated by Schlieren technique. For supersonic sprays, the concept of spray front angle is presented. Effects of Mach number of spray on the spray front angle are investigated. The results show that the shape of spray tip is similar to blunt body when fuel spray is at transonic region. If spray entered the supersonic region, the oblique shock waves are induced instead of normal shock wave. With the velocity of spray increasing, the spray front angle and shock wave angle are increased. The tip region of the supersonic fuel spray is commonly formed a cone. Mean droplet diameter of fuel spray is measured using Malvern’s Spraytec. Then the mean droplet diameter results are compared with three popular empirical models (Hiroyasu’s, Varde’s and Merrigton’s model). It is found that the Merrigton’s model shows a relative good correlation between models and experimental results. Finally, exponent of injection velocity in the Merrigton’s model is fitted with experimental results. PMID:28054555

  3. Single particle refuse-derived fuel devolatilization: Experimental measurements of reaction products

    SciTech Connect

    Lai, Weichuan; Krieger-Brockett, B. . Dept. of Chemical Engineering)

    1993-11-01

    The authors present experimentally measured devolatilization product yields from single particles of refuse-derived fuel (RDF), a more uniform, transportable municipal solid waste. Disposal costs and environmental concerns have stimulated interest in thermochemical conversion of this material to chemicals and fuels. The composition, reaction conditions, and particle properties were systematically varied over the range found in practice to develop quantitative measures that rank the process controllables' influence on altering the product slate. Specialized regression methods and experimental designs enhanced the accuracy in view of the feed heterogeneity and offer a general method to extract real effects from experimental and sample noise''. The results have been verified successfully using actual commercial RDF and fabricated compositions that surpass those normally found in municipal waste to anticipate the influence of trends in recycling. The results show that the reaction conditions have a greater influence on altering fuel utilization and the relative yields of char, condensibles, and gases than does the composition over the range found in MSW and RDF.

  4. Experimental investigation on the characteristics of supersonic fuel spray and configurations of induced shock waves

    NASA Astrophysics Data System (ADS)

    Wang, Yong; Yu, Yu-Song; Li, Guo-Xiu; Jia, Tao-Ming

    2017-01-01

    The macro characteristics and configurations of induced shock waves of the supersonic sprays are investigated by experimental methods. Visualization study of spray shape is carried out with the high-speed camera. The macro characteristics including spray tip penetration, velocity of spray tip and spray angle are analyzed. The configurations of shock waves are investigated by Schlieren technique. For supersonic sprays, the concept of spray front angle is presented. Effects of Mach number of spray on the spray front angle are investigated. The results show that the shape of spray tip is similar to blunt body when fuel spray is at transonic region. If spray entered the supersonic region, the oblique shock waves are induced instead of normal shock wave. With the velocity of spray increasing, the spray front angle and shock wave angle are increased. The tip region of the supersonic fuel spray is commonly formed a cone. Mean droplet diameter of fuel spray is measured using Malvern’s Spraytec. Then the mean droplet diameter results are compared with three popular empirical models (Hiroyasu’s, Varde’s and Merrigton’s model). It is found that the Merrigton’s model shows a relative good correlation between models and experimental results. Finally, exponent of injection velocity in the Merrigton’s model is fitted with experimental results.

  5. Trigger - and heat-transfer times measured during experimental molten-fuel-interactions

    SciTech Connect

    Spitznagel, N.; Dürig, T.; Zimanowski, B.

    2013-10-15

    A modified setup featuring high speed high resolution data and video recording was developed to obtain detailed information on trigger and heat transfer times during explosive molten fuel-coolant-interaction (MFCI). MFCI occurs predominantly in configurations where water is entrapped by hot melt. The setup was modified to allow direct observation of the trigger and explosion onset. In addition the influences of experimental control and data acquisition can now be more clearly distinguished from the pure phenomena. More precise experimental studies will facilitate the description of MFCI thermodynamics.

  6. Experimental analysis and semicontinuous simulation of low-temperature droplet evaporation of multicomponent fuels

    NASA Astrophysics Data System (ADS)

    Lehmann, S.; Lorenz, S.; Rivard, E.; Brüggemann, D.

    2015-01-01

    Low-pollutant and efficient combustion not only in internal combustion engines requires a balanced gaseous mixture of fuel and oxidizer. As fuels may contain several hundred different chemical species with different physicochemical properties as well as defined amounts of biogenic additives, e.g., ethanol, a thorough understanding of liquid fuel droplet evaporation processes is necessary to allow further engine optimization. We have studied the evaporation of fuel droplets at low ambient temperature. A non-uniform temperature distribution inside the droplet was already considered by including a finite thermal conductivity in a one-dimensional radial evaporation model (Rivard and Brüggemann in Chem Eng Sci 65(18):5137-5145, 2010). For a detailed analysis of droplet evaporation, two non-laser-based experimental setups have been developed. They allow a fast and relatively simple but yet precise measurement of diameter decrease and composition change. The first method is based on collecting droplets in a diameter range from 70 to 150 µm by a high-precision scale. A simultaneous evaluation of mass increase is employed for an accurate average diameter value determination. Subsequently, a gas chromatographic analysis of the collected droplets was conducted. In the second experiment, evaporation of even smaller droplets was optically analyzed by a high-speed shadowgraphy/schlieren microscope setup. A detailed analysis of evaporating E85 (ethanol/gasoline in a mass ratio of 85 %/15 %) and surrogate fuel droplets over a wide range of initial droplet diameters and ambient temperatures was conducted. The comparison of experimental and numerical results shows the applicability of the developed model over a large range of diameters and temperatures.

  7. Determination of Experimental Fuel Rod Parameters using 3D Modelling of PCMI with MPS Defect

    SciTech Connect

    Casagranda, Albert; Spencer, Benjamin Whiting; Pastore, Giovanni; Novascone, Stephen Rhead; Hales, Jason Dean; Williamson, Richard L; Martineau, Richard Charles

    2016-05-01

    An in-reactor experiment is being designed in order to validate the pellet-cladding mechanical interaction (PCMI) behavior of the BISON fuel performance code. The experimental parameters for the test rod being placed in the Halden Research Reactor are being determined using BISON simulations. The 3D model includes a missing pellet surface (MPS) defect to generate large local cladding deformations, which should be measureable after typical burnup times. The BISON fuel performance code is being developed at Idaho National Laboratory (INL) and is built on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. BISON supports both 2D and 3D finite elements and solves the fully coupled equations for solid mechanics, heat conduction and species diffusion. A number of fuel performance effects are included using models for swelling, densification, creep, relocation and fission gas production & release. In addition, the mechanical and thermal contact between the fuel and cladding is explicitly modelled using a master-slave based contact algorithm. In order to accurately predict PCMI effects, the BISON code includes the relevant physics involved and provides a scalable and robust solution procedure. The depth of the proposed MPS defect is being varied in the BISON model to establish an optimum value for the experiment. The experiment will be interrupted approximately every 6 months to measure cladding radial deformation and provide data to validate BISON. The complete rodlet (~20 discrete pellets) is being simulated using a 180° half symmetry 3D model with MPS defects at two axial locations. In addition, annular pellets will be used at the top and bottom of the pellet stack to allow thermocouples within the rod to measure the fuel centerline temperature. Simulation results will be presented to illustrate the expected PCMI behavior and support the chosen experimental design parameters.

  8. Analysis of Experimental Fuel Rod Parameters using 3D Modelling of PCMI with MPS Defect

    SciTech Connect

    Casagranda, Albert; Spencer, Benjamin Whiting; Pastore, Giovanni; Novascone, Stephen Rhead; Hales, Jason Dean; Williamson, Richard L; Martineau, Richard Charles

    2016-06-01

    An in-reactor experiment is being designed in order to validate the pellet-cladding mechanical interaction (PCMI) behavior of the BISON fuel performance code. The experimental parameters for the test rod being placed in the Halden Research Reactor are being determined using BISON simulations. The 3D model includes a missing pellet surface (MPS) defect to generate large local cladding deformations, which should be measureable after typical burnup times. The BISON fuel performance code is being developed at Idaho National Laboratory (INL) and is built on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. BISON supports both 2D and 3D finite elements and solves the fully coupled equations for solid mechanics, heat conduction and species diffusion. A number of fuel performance effects are included using models for swelling, densification, creep, relocation and fission gas production & release. In addition, the mechanical and thermal contact between the fuel and cladding is explicitly modelled using a master-slave based contact algorithm. In order to accurately predict PCMI effects, the BISON code includes the relevant physics involved and provides a scalable and robust solution procedure. The depth of the proposed MPS defect is being varied in the BISON model to establish an optimum value for the experiment. The experiment will be interrupted approximately every 6 months to measure cladding radial deformation and provide data to validate BISON. The complete rodlet (~20 discrete pellets) is being simulated using a 180° half symmetry 3D model with MPS defects at two axial locations. In addition, annular pellets will be used at the top and bottom of the pellet stack to allow thermocouples within the rod to measure the fuel centerline temperature. Simulation results will be presented to illustrate the expected PCMI behavior and support the chosen experimental design parameters.

  9. Untangling dopamine-adenosine receptor-receptor assembly in experimental parkinsonism in rats

    PubMed Central

    Fernández-Dueñas, Víctor; Taura, Jaume J.; Cottet, Martin; Gómez-Soler, Maricel; López-Cano, Marc; Ledent, Catherine; Watanabe, Masahiko; Trinquet, Eric; Pin, Jean-Philippe; Luján, Rafael; Durroux, Thierry; Ciruela, Francisco

    2015-01-01

    Parkinson’s disease (PD) is a dopaminergic-related pathology in which functioning of the basal ganglia is altered. It has been postulated that a direct receptor-receptor interaction – i.e. of dopamine D2 receptor (D2R) with adenosine A2A receptor (A2AR) (forming D2R-A2AR oligomers) – finely regulates this brain area. Accordingly, elucidating whether the pathology prompts changes to these complexes could provide valuable information for the design of new PD therapies. Here, we first resolved a long-standing question concerning whether D2R-A2AR assembly occurs in native tissue: by means of different complementary experimental approaches (i.e. immunoelectron microscopy, proximity ligation assay and TR-FRET), we unambiguously identified native D2R-A2AR oligomers in rat striatum. Subsequently, we determined that, under pathological conditions (i.e. in a rat PD model), D2R-A2AR interaction was impaired. Collectively, these results provide definitive evidence for alteration of native D2R-A2AR oligomers in experimental parkinsonism, thus conferring the rationale for appropriate oligomer-based PD treatments. PMID:25398851

  10. Leptospira Infection Interferes with the Prothrombinase Complex Assembly during Experimental Leptospirosis

    PubMed Central

    Vieira, Monica L.; de Andrade, Sonia A.; Morais, Zenaide M.; Vasconcellos, Silvio A.; Dagli, Maria Lucia Z.; Nascimento, Ana Lucia T. O.

    2017-01-01

    Leptospirosis is a worldwide zoonotic and neglected infectious disease of human and veterinary concern, caused by pathogenic Leptospira species. Although bleeding is a common symptom of severe leptospirosis, the cause of hemorrhage is not completely understood. In severe infections, modulation of hemostasis by pathogens is an important virulence mechanism, and hemostatic impairments such as coagulation/fibrinolysis dysfunction are frequently observed. Here, we analyze the coagulation status of experimentally infected hamsters in an attempt to determine coagulation interferences and the origin of leptospirosis hemorrhagic symptomatology. Hamsters were experimentally infected with L. interrogans. The lungs, kidneys, and livers were collected for culture, histopathology, and coagulation assays. L. interrogans infection disturbs normal coagulation in the organs of animals. Our results suggest the presence of a thrombin-like factor or FX activator, which is able to activate FII in the leptospirosis organ extracts. The activity of those factors is accelerated in the prothrombinase complex. Additionally, we show for the first time that live leptospires act as a surface for the prothrombinase complex assembly. Our results contribute to the understanding of leptospirosis pathophysiological mechanisms and may open new routes for the discovery of novel treatments in the severe manifestations of the disease.

  11. Performance of polymer nano composite membrane electrode assembly using Alginate as a dopant in polymer electrolyte membrane fuel cell

    NASA Astrophysics Data System (ADS)

    Mulijani, S.

    2016-11-01

    Polymer membrane and composite polymer for membrane electrode assembly (MEAs) are synthesized and studied for usage in direct methanol fuel cell (DMFC). In this study, we prepared 3 type of MEAs, polystyrene (PS), sulfonated polystyrene (SPS) and composite polymer SPS-alginat membrane via catalyst hot pressed method. The performance and properties of prepared MEAs were evaluated and analyzed by impedance spectrometry and scanning electron microscopy (SEM). The result showed that, water up take of MEA composite polymer SPS-alginate was obtained higher than that in SPS and PS. The proton conductivity of MEA-SPS-alginate was also higher than that PS and PSS. SEM characterization revealed that the intimate contact between the carbon catalyst layers (CL) and the membranes, and the uniformly porous structure correlate positively with the MEAs prepared by hot pressed method, exhibiting high performances for DMFC.

  12. Performance of polymer nano composite membrane electrode assembly using Alginate as a dopant in polymer electrolyte membrane fuel cell

    NASA Astrophysics Data System (ADS)

    Mulijani, S.

    2017-01-01

    Polymer membrane and composite polymer for membrane electrode assembly (MEAs) are synthesized and studied for usage in direct methanol fuel cell (DMFC). In this study, we prepared 3 type of MEAs, polystyrene (PS), sulfonated polystyrene (SPS) and composite polymer SPS-alginat membrane via catalyst hot pressed method. The performance and properties of prepared MEAs were evaluated and analyzed by impedance spectrometry and scanning electron microscopy (SEM). The result showed that, water up take of MEA composite polymer SPS-alginate was obtained higher than that in SPS and PS. The proton conductivity of MEA-SPS-alginate was also higher than that PS and PSS. SEM characterization revealed that the intimate contact between the carbon catalyst layers (CL) and the membranes, and the uniformly porous structure correlate positively with the MEAs prepared by hot pressed method, exhibiting high performances for DMFC.

  13. Durability of Membrane Electrode Assemblies (MEAs) in PEM Fuel Cells Operated on Pure Hydrogen and Oxygen

    NASA Technical Reports Server (NTRS)

    Stanic, Vesna; Braun, James; Hoberecht, Mark

    2003-01-01

    Proton exchange membrane (PEM) fuel cells are energy sources that have the potential to replace alkaline fuel cells for space programs. Broad power ranges, high peak-to-nominal power capabilities, low maintenance costs, and the promise of increased life are the major advantages of PEM technology in comparison to alkaline technology. The probability of PEM fuel cells replacing alkaline fuel cells for space applications will increase if the promise of increased life is verified by achieving a minimum of 10,000 hours of operating life. Durability plays an important role in the process of evaluation and selection of MEAs for Teledyne s Phase I contract with the NASA Glenn Research Center entitled Proton Exchange Membrane Fuel cell (PEMFC) Power Plant Technology Development for 2nd Generation Reusable Launch Vehicles (RLVs). For this contract, MEAs that are typically used for H2/air operation were selected as potential candidates for H2/O2 PEM fuel cells because their catalysts have properties suitable for O2 operation. They were purchased from several well-established MEA manufacturers who are world leaders in the manufacturing of diverse products and have committed extensive resources in an attempt to develop and fully commercialize MEA technology. A total of twelve MEAs used in H2/air operation were initially identified from these manufacturers. Based on the manufacturers specifications, nine of these were selected for evaluation. Since 10,000 hours is almost equivalent to 14 months, it was not possible to perform continuous testing with each MEA selected during Phase I of the contract. Because of the lack of time, a screening test on each MEA was performed for 400 hours under accelerated test conditions. The major criterion for an MEA pass or fail of the screening test was the gas crossover rate. If the gas crossover rate was higher than the membrane intrinsic permeability after 400 hours of testing, it was considered that the MEA had failed the test. Three types of

  14. Assembly of a Cost-Effective Anode Using Palladium Nanoparticles for Alkaline Fuel Cell Applications

    PubMed Central

    2015-01-01

    Nanotechnology allows the synthesis of nanoscale catalysts, which offer an efficient alternative for fuel cell applications. In this laboratory experiment, the student selects a cost-effective anode for fuel cells by comparing three different working electrodes. These are commercially available palladium (Pd) and glassy carbon (GC) electrodes, and a carbon paste (CP) electrode that is prepared by the students in the laboratory. The GC and CP were modified with palladium nanoparticles (PdNP) suspensions. The electrodes efficiencies were studied for ethanol oxidation in alkaline solution using cyclic voltammetry techniques. The ethanol oxidation currents obtained were used to determine the current density using the geometric and surface area of each electrode. Finally, students were able to choose the best electrode and relate catalytic activity to surface area for ethanol oxidation in alkaline solution by completing a critical analysis of the cyclic voltammetry results. With this activity, fundamental electrochemical concepts were reinforced. PMID:25691801

  15. Nano composite membrane-electrode assembly formation for fuel cell-modeling aspects

    NASA Astrophysics Data System (ADS)

    Vaivars, G.; Linkov, V.

    2007-12-01

    Long term stability is an essential requirement for fuel cell applications in automobile and stationary energy systems. In these systems the agglomeration of the catalyst nanoparticles is a well-known phenomenon which cannot be easily overcome or compensated for by re-designing the system. A direct result of this occurrence is the irreversible decrease of the electrochemical performance. Irregularities in electric field distribution are one root cause for migration and subsequent agglomeration of the catalyst nanoparticle. In this work, the impact of the electrode mechanical deformation on electric field distribution was studied using a computer modeling approach. Model of a Proton Exchange Membrane (PEM) fuel cell with interdigitated flow field from Comsol Chemical Engineering/Electrochemical Engineering Module library was used for simulations. It was established that by minimizing the backing layer deformation it is possible to achieve some improvement in current distribution.

  16. Assembly of a Cost-Effective Anode Using Palladium Nanoparticles for Alkaline Fuel Cell Applications.

    PubMed

    Feliciano-Ramos, Ileana; Casañas-Montes, Barbara; García-Maldonado, María M; Menéndez, Christian L; Mayol, Ana R; Díaz-Vázquez, Liz M; Cabrera, Carlos R

    2015-02-10

    Nanotechnology allows the synthesis of nanoscale catalysts, which offer an efficient alternative for fuel cell applications. In this laboratory experiment, the student selects a cost-effective anode for fuel cells by comparing three different working electrodes. These are commercially available palladium (Pd) and glassy carbon (GC) electrodes, and a carbon paste (CP) electrode that is prepared by the students in the laboratory. The GC and CP were modified with palladium nanoparticles (PdNP) suspensions. The electrodes efficiencies were studied for ethanol oxidation in alkaline solution using cyclic voltammetry techniques. The ethanol oxidation currents obtained were used to determine the current density using the geometric and surface area of each electrode. Finally, students were able to choose the best electrode and relate catalytic activity to surface area for ethanol oxidation in alkaline solution by completing a critical analysis of the cyclic voltammetry results. With this activity, fundamental electrochemical concepts were reinforced.

  17. Development of self-interrogation neutron resonance densitometry (SINRD) to measure U-235 and Pu-239 content in a PWR spent fuel assembly

    SciTech Connect

    Lafleur, Adrienne M; Charlton, William S; Menlove, Howard O; Swinhoe, Martyn T

    2009-01-01

    The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U and {sup 239}Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the {sup 244}Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using {sup 235}U and {sup 239}Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

  18. Moments applied in the manual assembly of space structures - Ease biomechanics results from STS-61B. [Experimental Assembly of Structures in EVA

    NASA Technical Reports Server (NTRS)

    Cousins, D.; Akin, D. L.

    1989-01-01

    Measurements of the level and pattern of moments applied in the manual assembly of a space structure were made in extravehicular activity (EVA) and neutral buoyancy simulation (NBS). The Experimental Assembly of Structures in EVA program included the repeated assembly of a 3.6 m tetrahedral truss structure in EVA on STS-61B after extensive neutral buoyancy crew training. The flight and training structures were of equivalent mass and geometry to allow a direct correlation between EVA and NBS performance. A stereo photographic motion camera system was used to reconstruct in three dimensions rotational movements of structural beams during assembly. Moments applied in these manual handling tasks were calculated on the basis of the reconstructed movements taking into account effects of inertia, drag and virtual mass. Applied moments of 2.0 Nm were typical for beam rotations in EVA. Corresponding applied moments in NBS were typically up to five times greater. Moments were applied as impulses separated by several seconds of coasting in both EVA and NBS. Decelerating impulses were only infrequently observed in NBS.

  19. Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    SciTech Connect

    Favalli, Andrea; Vo, D.; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Schwalbach, P.; Sjoland, A.; Tobin, Stephen J.; Trellue, Holly; Vaccaro, S.

    2016-02-26

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  20. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    NASA Astrophysics Data System (ADS)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  1. Experience with the loading and transport of fuel assembly transport casks, including CASTOR casks, and the radiation exposure of personnel.

    PubMed

    Bentele, W; Kinzelmann, T

    1999-12-01

    In 1997 and 1998, six spent fuel assembly transports started from the nuclear power plant Gemeinschaftskernkraftwerk Neckar (GKN), using CASTOR-V19 casks. Professor Kuni of Marburg University challenged the statement made by the German Federal Office for Radiation Protection (Bundesamt für Strahlenschutz (BfS)) based on accepted scientific knowledge, according to which so-called CASTOR transports present no risk, either to the population or to the escorting police units. This paper shows that the collective dose during the loading of the CASTOR casks amounted to 4.5 mSv (gamma and neutrons) per cask at the most, and that the maximum individual dose amounted to 0.26 mSv. In addition to these doses, the collective dose during handling and transport must be considered: this amounted to 0.35 mSv (gamma and neutrons). The dose to the police escort was <0.1 mSv (gamma and neutrons). In the light of these circumstances, this report is presented on contamination determined during the transport of CASTOR casks and of other spent fuel casks. The controls of spent fuel transports carried out since 1978, mainly with NTL 11 spent fuel casks, revealed that about one fifth of the transport casks which left the GKN with a surface contamination of <4 Bq cm(-2) (limit for surface contamination), presented degrees of contamination >4 Bq cm(-2) upon reaching the Valognes/Cogema terminal. However, transport casks coming from French plants also revealed degrees of contamination >4 Bq cm(-2), as well as 'hot spots'. No such contamination was found on NTL 11 casks transported from the GKN to Sellafield. Neither was any increased contamination found upon the arrival of CASTOR-V19 casks transported from GKN to Gorleben or Ahaus. The partially sensationalist media reports were inversely proportional to the actual radiological relevance of the matter. The German Commission on Radiation Protection (SSK) confirmed that the radiological effect of such contaminated spent fuel transports is

  2. Sobol's sensitivity analysis for a fuel cell stack assembly model with the aid of structure-selection techniques

    NASA Astrophysics Data System (ADS)

    Zhang, Wei; Cho, Chongdu; Piao, Changhao; Choi, Hojoon

    2016-01-01

    This paper presents a novel method for identifying the main parameters affecting the stress distribution of the components used in assembly modeling of proton exchange membrane fuel cell (PEMFC) stack. This method is a combination of an approximation model and Sobol's method, which allows a fast global sensitivity analysis for a set of uncertain parameters using only a limited number of calculations. Seven major parameters, i.e., Young's modulus of the end plate and the membrane electrode assembly (MEA), the contact stiffness between the MEA and bipolar plate (BPP), the X and Y positions of the bolts, the pressure of each bolt, and the thickness of the end plate, are investigated regarding their effect on four metrics, i.e., the maximum stresses of the MEA, BPP, and end plate, and the stress distribution percentage of the MEA. The analysis reveals the individual effects of each parameter and its interactions with the other parameters. The results show that the X position of a bolt has a major influence on the maximum stresses of the BPP and end plate, whereas the thickness of the end plate has the strongest effect on both the maximum stress and the stress distribution percentage of the MEA.

  3. Analysis of burnup and isotopic compositions of BWR 9 x 9 UO{sub 2} fuel assemblies

    SciTech Connect

    Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.

    2012-07-01

    In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO{sub 2} fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for {sup 238}Pu, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148}Nd, {sup 134}Cs, {sup 154}Eu, {sup 152}Sm, {sup 154}Gd, and {sup 157}Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

  4. A membrane-less enzymatic fuel cell with layer-by-layer assembly of redox polymer and enzyme over graphite electrodes.

    PubMed

    Rengaraj, Saravanan; Mani, Vigneshwaran; Kavanagh, Paul; Rusling, James; Leech, Dónal

    2011-11-21

    Layer-by-layer (LBL) assembly of alternate osmium redox polymers and glucose oxidase, at anode, and laccase, at cathode, using graphite electrodes form a membrane-less glucose/O(2) enzymatic fuel cell providing a power density of 103 μW cm(-2) at pH 5.5.

  5. The problems of mass transfer and formation of deposits of corrosion products on fuel assemblies of a VVER-1200 reactor

    NASA Astrophysics Data System (ADS)

    Rodionov, Yu. A.; Kritskii, V. G.; Berezina, I. G.; Gavrilov, A. V.

    2014-03-01

    On the basis of examination of materials published both in Russia and abroad, as well as their own investigations, the authors explain the reasons for the occurrence of such effects as AOA (Axial Offset Anomalies) and an increase in the coolant pressure difference in the core of nuclear reactors of the VVER type. To detect the occurrence of the AOA effect, the authors suggest using the specific activity of 58Co in the coolant. In the VVER-1200 design the thermohydraulic regime for fuel assemblies in the first year of their service life involves slight boiling of the coolant in the upper part of the core, which may induce the occurrence of the AOA effect, intensification of corrosion of fuel claddings, and abnormal increase in deposition of corrosion products. Radiolysis of the water coolant in the boiling section (boiling in pores of deposits) may intensify not only general corrosion but also a localized (nodular) one. As a result of intensification of the corrosion processes and growth of deposits, deterioration of the radiation situation in the rooms of the primary circuit of a VVER-1200 reactor as compared to that at nuclear power plants equipped with reactors of the VVER-1000 type is possible. Recommendations for preventing the AOA effect at nuclear power plants with VVER-1200 reactors on the matter of the direction of further investigations are made.

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

    SciTech Connect

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

    1995-12-31

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

  7. Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

    SciTech Connect

    Bardet, Philippe; Ricciardi, Guillaume

    2016-01-31

    Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWR fuel bundle behavior during seismic transients.

  8. Experimental and Computational Study on the Cusp-DEC and TWDEC for Advanced Fueled Fusion

    SciTech Connect

    Tomita, Y.; Yasaka, Y.; Takeno, H.; Ishikawa, M.; Nemoto, T.

    2005-01-15

    Experimental and computational results of direct energy converters (DECs) for advanced fueled fusion such as D-{sup 3}He are presented. Kinetic energy of thermal component of end loss plasma is converted to electricity by using the Cusp DEC. The proof-of-principle experiments of a single slanted cusp have been carried out and verified the faculty of the configuration. To improve a separation of electrons from ions, numerical simulation shows a Helmholtz magnetic configuration with a uniform magnetic field is more effective than the Cusp DEC. The fusion-produced high-energy ions like 15 MeV protons in D-{sup 3}He fueled fusion can pass through the Cusp DEC without disturbing their orbits and enter a traveling-wave direct energy converter (TWDEC). Small scale experiments have shown the effectiveness of the TWDEC and the numerical simulation on optimization of interval of electrodes in a decelerator gives high conversion efficiency up to 60 %.

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

    NASA Technical Reports Server (NTRS)

    Phillips, W. M.

    1973-01-01

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

  10. Radiation Damage in Nuclear Fuel for Advanced Burner Reactors: Modeling and Experimental Validation

    SciTech Connect

    Jensen, Niels Gronbech; Asta, Mark; Ozolins, Nigel Browning'Vidvuds; de Walle, Axel van; Wolverton, Christopher

    2011-12-29

    The consortium has completed its existence and we are here highlighting work and accomplishments. As outlined in the proposal, the objective of the work was to advance the theoretical understanding of advanced nuclear fuel materials (oxides) toward a comprehensive modeling strategy that incorporates the different relevant scales involved in radiation damage in oxide fuels. Approaching this we set out to investigate and develop a set of directions: 1) Fission fragment and ion trajectory studies through advanced molecular dynamics methods that allow for statistical multi-scale simulations. This work also includes an investigation of appropriate interatomic force fields useful for the energetic multi-scale phenomena of high energy collisions; 2) Studies of defect and gas bubble formation through electronic structure and Monte Carlo simulations; and 3) an experimental component for the characterization of materials such that comparisons can be obtained between theory and experiment.

  11. Dynamic modeling, experimental evaluation, optimal design and control of integrated fuel cell system and hybrid energy systems for building demands

    NASA Astrophysics Data System (ADS)

    Nguyen, Gia Luong Huu

    Fuel cells can produce electricity with high efficiency, low pollutants, and low noise. With the advent of fuel cell technologies, fuel cell systems have since been demonstrated as reliable power generators with power outputs from a few watts to a few megawatts. With proper equipment, fuel cell systems can produce heating and cooling, thus increased its overall efficiency. To increase the acceptance from electrical utilities and building owners, fuel cell systems must operate more dynamically and integrate well with renewable energy resources. This research studies the dynamic performance of fuel cells and the integration of fuel cells with other equipment in three levels: (i) the fuel cell stack operating on hydrogen and reformate gases, (ii) the fuel cell system consisting of a fuel reformer, a fuel cell stack, and a heat recovery unit, and (iii) the hybrid energy system consisting of photovoltaic panels, fuel cell system, and energy storage. In the first part, this research studied the steady-state and dynamic performance of a high temperature PEM fuel cell stack. Collaborators at Aalborg University (Aalborg, Denmark) conducted experiments on a high temperature PEM fuel cell short stack at steady-state and transients. Along with the experimental activities, this research developed a first-principles dynamic model of a fuel cell stack. The dynamic model developed in this research was compared to the experimental results when operating on different reformate concentrations. Finally, the dynamic performance of the fuel cell stack for a rapid increase and rapid decrease in power was evaluated. The dynamic model well predicted the performance of the well-performing cells in the experimental fuel cell stack. The second part of the research studied the dynamic response of a high temperature PEM fuel cell system consisting of a fuel reformer, a fuel cell stack, and a heat recovery unit with high thermal integration. After verifying the model performance with the

  12. Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    DOE PAGES

    Favalli, Andrea; Vo, D.; Grogan, Brandon R.; ...

    2016-02-26

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

  13. Validation of fuel bundle mechanical performance code ETOILE with bundle/duct interaction experimental data

    SciTech Connect

    Nakagawa, Masatoshi )

    1993-04-01

    Validation of the ETOILE code through a comparison with experimental bundle/duct interaction (BDI) data is discussed. ETOILE is a newly developed three-dimensional finite element program that uses a new analytical method to predict distortions and mechanical behavior in wire-wrapped-type fuel-pin bundles during irradiation in liquid-metal fast breeder reactor cores. Comparisons between the ETOILE solutions and the experimental data for bundle stiffnesses and minimum pin-to-pin and pin-to-duct clearances under bundle compression suggest that BDI performance can be predicted reasonably well with a suitable choice of friction coefficient and initial spiral wire displacement. Application of the code in the analysis of the mechanical behavior of soft bundles with distributed wireless pins is also presented to demonstrate the effectiveness of this design in reducing the interaction forces between a fuel-pin bundle and a duct wall under bundle compression. Agreement with the experimental data is fairly good for the reduction in bundle stiffness when the configuration is changed from the normal bundle to the soft bundle.

  14. A robust fuel cell cathode catalyst assembled with nitrogen-doped carbon nanohorn and platinum nanoclusters

    NASA Astrophysics Data System (ADS)

    Zhang, Linwei; Zheng, Ning; Gao, Ang; Zhu, Chunmei; Wang, Zhiyong; Wang, Yuan; Shi, Zujin; Liu, Yan

    2012-12-01

    A highly durable and active nanocomposite cathode catalyst (Pt/NSWCNH) is assembled with “unprotected” Pt nanoclusters and nitrogen-doped single-wall carbon nanohorns (NSWCNH) as building blocks by a convenient process. The specific catalytic activity and mass catalytic activity for the oxygen reduction reaction over Pt/NSWCNH is 1.60 and 1.75 times as high as those over a commercial Pt/C catalyst, respectively. There is no obvious loss in the catalytic activity of Pt/NSWCNH after potential cycling from 0.6 to 1.1 V versus RHE for 15,000 cycles at 30 °C, under the oxidizing conditions for the electrochemically catalytic reduction of O2. TEM characterization results reveal that, during the accelerated aging tests, Pt nanoparticles in Pt/NSWCNH are more stable than that in Pt/C-JM, showing a low increase in the particle size.

  15. Electrode assembly for use in a solid polymer electrolyte fuel cell

    DOEpatents

    Raistrick, Ian D.

    1989-01-01

    A gas reaction fuel cell may be provided with a solid polymer electrolyte membrane. Porous gas diffusion electrodes are formed of carbon particles supporting a catalyst which is effective to enhance the gas reactions. The carbon particles define interstitial spaces exposing the catalyst on a large surface area of the carbon particles. A proton conducting material, such as a perfluorocarbon copolymer or ruthenium dioxide contacts the surface areas of the carbon particles adjacent the interstitial spaces. The proton conducting material enables protons produced by the gas reactions adjacent the supported catalyst to have a conductive path with the electrolyte membrane. The carbon particles provide a conductive path for electrons. A suitable electrode may be formed by dispersing a solution containing a proton conducting material over the surface of the electrode in a manner effective to coat carbon surfaces adjacent the interstitial spaces without impeding gas flow into the interstitial spaces.

  16. Mechanism of Pinhole Formation in Membrane Electrode Assemblies for PEM Fuel Cells

    NASA Technical Reports Server (NTRS)

    Stanic, Vesna; Hoberecht, Mark

    2004-01-01

    The pinhole formation mechanism was studied with a variety of MEAs using ex-situ and in-situ methods. The ex-situ tests included the MEA aging in oxygen and MEA heat of ignition. In-situ durability tests were performed in fuel cells at different operating conditions with hydrogen and oxygen. After the in-situ failure, MEAs were analyzed with an Olympus BX 60 optical microscope and Cambridge 120 scanning electron microscope. MEA chemical analysis was performed with an IXRF EDS microanalysis system. The MEA failure analyses showed that pinholes and tears were the MEA failure modes. The pinholes appeared in MEA areas where the membrane thickness was drastically reduced. Their location coincided with the stress concentration points, indicating that membrane creep was responsible for their formation. Some of the pinholes detected had contaminant particles precipitated within the membrane. This mechanism of pinhole formation was correlated to the polymer blistering.

  17. Assembly of high-areal-density deuterium-tritium fuel from indirectly driven cryogenic implosions.

    PubMed

    Mackinnon, A J; Kline, J L; Dixit, S N; Glenzer, S H; Edwards, M J; Callahan, D A; Meezan, N B; Haan, S W; Kilkenny, J D; Döppner, T; Farley, D R; Moody, J D; Ralph, J E; MacGowan, B J; Landen, O L; Robey, H F; Boehly, T R; Celliers, P M; Eggert, J H; Krauter, K; Frieders, G; Ross, G F; Hicks, D G; Olson, R E; Weber, S V; Spears, B K; Salmonsen, J D; Michel, P; Divol, L; Hammel, B; Thomas, C A; Clark, D S; Jones, O S; Springer, P T; Cerjan, C J; Collins, G W; Glebov, V Y; Knauer, J P; Sangster, C; Stoeckl, C; McKenty, P; McNaney, J M; Leeper, R J; Ruiz, C L; Cooper, G W; Nelson, A G; Chandler, G G A; Hahn, K D; Moran, M J; Schneider, M B; Palmer, N E; Bionta, R M; Hartouni, E P; LePape, S; Patel, P K; Izumi, N; Tommasini, R; Bond, E J; Caggiano, J A; Hatarik, R; Grim, G P; Merrill, F E; Fittinghoff, D N; Guler, N; Drury, O; Wilson, D C; Herrmann, H W; Stoeffl, W; Casey, D T; Johnson, M G; Frenje, J A; Petrasso, R D; Zylestra, A; Rinderknecht, H; Kalantar, D H; Dzenitis, J M; Di Nicola, P; Eder, D C; Courdin, W H; Gururangan, G; Burkhart, S C; Friedrich, S; Blueuel, D L; Bernstein, L A; Eckart, M J; Munro, D H; Hatchett, S P; Macphee, A G; Edgell, D H; Bradley, D K; Bell, P M; Glenn, S M; Simanovskaia, N; Barrios, M A; Benedetti, R; Kyrala, G A; Town, R P J; Dewald, E L; Milovich, J L; Widmann, K; Moore, A S; LaCaille, G; Regan, S P; Suter, L J; Felker, B; Ashabranner, R C; Jackson, M C; Prasad, R; Richardson, M J; Kohut, T R; Datte, P S; Krauter, G W; Klingman, J J; Burr, R F; Land, T A; Hermann, M R; Latray, D A; Saunders, R L; Weaver, S; Cohen, S J; Berzins, L; Brass, S G; Palma, E S; Lowe-Webb, R R; McHalle, G N; Arnold, P A; Lagin, L J; Marshall, C D; Brunton, G K; Mathisen, D G; Wood, R D; Cox, J R; Ehrlich, R B; Knittel, K M; Bowers, M W; Zacharias, R A; Young, B K; Holder, J P; Kimbrough, J R; Ma, T; La Fortune, K N; Widmayer, C C; Shaw, M J; Erbert, G V; Jancaitis, K S; DiNicola, J M; Orth, C; Heestand, G; Kirkwood, R; Haynam, C; Wegner, P J; Whitman, P K; Hamza, A; Dzenitis, E G; Wallace, R J; Bhandarkar, S D; Parham, T G; Dylla-Spears, R; Mapoles, E R; Kozioziemski, B J; Sater, J D; Walters, C F; Haid, B J; Fair, J; Nikroo, A; Giraldez, E; Moreno, K; Vanwonterghem, B; Kauffman, R L; Batha, S; Larson, D W; Fortner, R J; Schneider, D H; Lindl, J D; Patterson, R W; Atherton, L J; Moses, E I

    2012-05-25

    The National Ignition Facility has been used to compress deuterium-tritium to an average areal density of ~1.0±0.1 g cm(-2), which is 67% of the ignition requirement. These conditions were obtained using 192 laser beams with total energy of 1-1.6 MJ and peak power up to 420 TW to create a hohlraum drive with a shaped power profile, peaking at a soft x-ray radiation temperature of 275-300 eV. This pulse delivered a series of shocks that compressed a capsule containing cryogenic deuterium-tritium to a radius of 25-35 μm. Neutron images of the implosion were used to estimate a fuel density of 500-800 g cm(-3).

  18. Microscopic Fuel Particles Produced by Self-Assembly of Actinide Nanoclusters on Carbon Nanomaterials

    SciTech Connect

    Na, Chongzheng

    2016-10-17

    Many consider further development of nuclear power to be essential for sustained development of society; however, the fuel forms currently used are expensive to recycle. In this project, we sought to create the knowledge and knowhow that are needed to produce nanocomposite materials by directly depositing uranium nanoclusters on networks of carbon-­ based nanomaterials. The objectives of the proposed work were to (1) determine the control of uranium nanocluster surface chemistry on nanocomposite formation, (2) determine the control of carbon nanomaterial surface chemistry on nanocomposite formation, and (3) develop protocols for synthesizing uranium-­carbon nanomaterials. After examining a wide variety of synthetic methods, we show that synthesizing graphene-­supported UO2 nanocrystals in polar ethylene glycol compounds by polyol reduction under boiling reflux can enable the use of an inexpensive graphene precursor graphene oxide in the production of uranium-carbon nanocomposites in a one-­pot process. We further show that triethylene glycol is the most suitable solvent for producing nanometer-­sized UO2 crystals compared to monoethylene glycol, diethylene glycol, and polyethylene glycol. Graphene-­supported UO2 nanocrystals synthesized with triethylene glycol show evidence of heteroepitaxy, which can be beneficial for facilitating heat transfer in nuclear fuel particles. Furthermore, we show that graphene-supported UO2 nanocrystals synthesized by polyol reduction can be readily stored in alcohols, preventing oxidation from the prevalent oxygen in air. Together, these methods provide a facile approach for preparing and storing graphene-supported UO nanocrystals for further investigation and development under ambient conditions.

  19. Experimental and theoretical investigation of water removal from DMAZ liquid fuel by an adsorption process

    NASA Astrophysics Data System (ADS)

    Ghanbari, Shahram; Vaferi, Behzad

    2015-07-01

    2-dimethylaminoethylazide (DMAZ) is a new liquid fuel that has made significant progress in bio/mono propellant rocket engines in recent years. Purification of DMAZ fuel by reducing its water content using various adsorbents including zeolites, calcium chloride and nano-particles is experimentally and theoretically investigated. The highest water adsorption of 92.6% from the DMAZ solution is obtained by the CaCl2 adsorbent within 10 min. Four different artificial neural networks (ANN) are examined to correlate an extent of removed water from the DMAZ solution to its affecting parameters. The performed regression analysis indicated that water initial concentration (WIC), adsorbent types, solution temperature, contact time and adsorbent dosage are the most important affecting variables for water sorption from the DMAZ solution. The accomplished statistical analysis demonstrated a multi-layer perceptron neural network (MLPNN) with seven hidden neurons and is the most accurate approach for modeling the considered task. The obtained results showed that the proposed MLPNN model could be successfully employed for accurate prediction of an amount of water removal from the DMAZ fuel solution by the adsorption process.

  20. Experimental verification of the thermodynamic properties for a jet-A fuel

    NASA Technical Reports Server (NTRS)

    Graciasalcedo, Carmen M.; Brabbs, Theodore A.; Mcbride, Bonnie J.

    1988-01-01

    Thermodynamic properties for a Jet-A fuel were determined by Shell Development Company in 1970 under a contract for NASA Lewis Research Center. The polynomial fit necessary to include Jet-A fuel (liquid and gaseous phases) in the library of thermodynamic properties of the NASA Lewis Chemical Equilibrium Program is calculated. To verify the thermodynamic data, the temperatures of mixtures of liquid Jet-A injected into a hot nitrogen stream were experimentally measured and compared to those calculated by the program. Iso-octane, a fuel for which the thermodynamic properties are well known, was used as a standard to calibrate the apparatus. The measured temperatures for the iso-octane/nitrogen mixtures reproduced the calculated temperatures except for a small loss due to the non-adiabatic behavior of the apparatus. The measurements for Jet-A were corrected for this heat loss and showed excellent agreement with the calculated temperatures. These experiments show that this process can be adequately described by the thermodynamic properties fitted for the Chemical Equilibrium Program.

  1. An Artificial Gravity Spacecraft Approach which Minimizes Mass, Fuel and Orbital Assembly Reg

    NASA Astrophysics Data System (ADS)

    Bell, L.

    2002-01-01

    The Sasakawa International Center for Space Architecture (SICSA) is undertaking a multi-year research and design study that is exploring near and long-term commercial space development opportunities. Space tourism in low-Earth orbit (LEO), and possibly beyond LEO, comprises one business element of this plan. Supported by a financial gift from the owner of a national U.S. hotel chain, SICSA has examined opportunities, requirements and facility concepts to accommodate up to 100 private citizens and crewmembers in LEO, as well as on lunar/planetary rendezvous voyages. SICSA's artificial gravity Science Excursion Vehicle ("AGSEV") design which is featured in this presentation was conceived as an option for consideration to enable round-trip travel to Moon and Mars orbits and back from LEO. During the course of its development, the AGSEV would also serve other important purposes. An early assembly stage would provide an orbital science and technology testbed for artificial gravity demonstration experiments. An ultimate mature stage application would carry crews of up to 12 people on Mars rendezvous missions, consuming approximately the same propellant mass required for lunar excursions. Since artificial gravity spacecraft that rotate to create centripetal accelerations must have long spin radii to limit adverse effects of Coriolis forces upon inhabitants, SICSA's AGSEV design embodies a unique tethered body concept which is highly efficient in terms of structural mass and on-orbit assembly requirements. The design also incorporates "inflatable" as well as "hard" habitat modules to optimize internal volume/mass relationships. Other important considerations and features include: maximizing safety through element and system redundancy; means to avoid destabilizing mass imbalances throughout all construction and operational stages; optimizing ease of on-orbit servicing between missions; and maximizing comfort and performance through careful attention to human needs. A

  2. Advanced computational tools for PEM fuel cell design. Part 2. Detailed experimental validation and parametric study

    NASA Astrophysics Data System (ADS)

    Sui, P. C.; Kumar, S.; Djilali, N.

    This paper reports on the systematic experimental validation of a comprehensive 3D CFD-based computational model presented and documented in Part 1. Simulations for unit cells with straight channels, similar to the Ballard Mk902 hardware, are performed and analyzed in conjunction with detailed current mapping measurements and water mass distributions in the membrane-electrode assembly. The experiments were designed to display sensitivity of the cell over a range of operating parameters including current density, humidification, and coolant temperature, making the data particularly well suited for systematic validation. Based on the validation and analysis of the predictions, values of model parameters, including the electro-osmotic drag coefficient, capillary diffusion coefficient, and catalyst specific surface area are determined adjusted to fit experimental data of current density and MEA water content. The predicted net water flux out of the anode (normalized by the total water generated) increases as anode humidification water flow rate is increased, in agreement with experimental results. A modification of the constitutive equation for the capillary diffusivity of water in the porous electrodes that attempts to incorporate the experimentally observed immobile (or irreducible) saturation yields a better fit of the predicted MEA water mass with experimental data. The specific surface area parameter used in the catalyst layer model is found to be effective in tuning the simulations to predict the correct cell voltage over a range of stoichiometries.

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

    SciTech Connect

    Samuel J. Miller; Hakan Ozaltun

    2012-11-01

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

  4. An Experimental Investigation of Hypergolic Ignition Delay of Hydrogen Peroxide with Fuel Mixtures

    NASA Technical Reports Server (NTRS)

    Blevins, John A.; Gostowski, Rudy; Chianese, Silvio

    2003-01-01

    An experimental investigation of hypergolicity and ignition delay of fuel mixtures with hydrogen peroxide is presented. Example results of high speed photography and schleiren from drop tests are shown. Also, a discussion of the sensitivity to experimental parameters such as drop size and subsequent uncertainty considerations of ignition delay results is presented. It is shown that using the described setup on the mixtures presented, the precision uncertainty is on the order of 6% of average ignition delay and 5% of average decomposition delay. This represents sufficient repeatability for first order discrimination of ignition delay for propellant development and screening. Two mixtures, each using commonly available amines and transition metal compounds, are presented as examples that result in ignition delays on the order of 10 milliseconds.

  5. Understanding electricity generation in osmotic microbial fuel cells through integrated experimental investigation and mathematical modeling.

    PubMed

    Qin, Mohan; Ping, Qingyun; Lu, Yaobin; Abu-Reesh, Ibrahim M; He, Zhen

    2015-11-01

    Osmotic microbial fuel cells (OsMFCs) are a new type of MFCs with integrating forward osmosis (FO). However, it is not well understood why electricity generation is improved in OsMFCs compared to regular MFCs. Herein, an approach integrating experimental investigation and mathematical model was adopted to address the question. Both an OsMFC and an MFC achieved similar organic removal efficiency, but the OsMFC generated higher current than the MFC with or without water flux, resulting from the lower resistance of FO membrane. Combining NaCl and glucose as a catholyte demonstrated that the catholyte conductivity affected the electricity generation in the OsMFC. A mathematical model of OsMFCs was developed and validated with the experimental data. The model predicated the variation of internal resistance with increasing water flux, and confirmed the importance of membrane resistance. Increasing water flux with higher catholyte conductivity could decrease the membrane resistance.

  6. Experimental and numerical studies of burning velocities and kinetic modeling for practical and surrogate fuels

    NASA Astrophysics Data System (ADS)

    Zhao, Zhenwei

    To help understand the fuel oxidation process in practical combustion environments, laminar flame speeds and high temperature chemical kinetic models were studied for several practical fuels and "surrogate" fuels, such as propane, dimethyl ether (DME), and primary reference fuel (PRF) mixtures, gasoline and n-decane. The PIV system developed for the present work is described. The general principles for PIV measurements are outlined and the specific considerations are also reported. Laminar flame speeds were determined for propane/air over a range of equivalence ratios at initial temperature of 298 K, 500 K and 650 K and atmospheric pressure. Several data sets for propane/air laminar flame speeds with N 2 dilution are also reported. These results are compared to the literature data collected at the same conditions. The propane flame speed is also numerically calculated with a detailed kinetic model and multi component diffusion, including Soret effects. This thesis also presents experimentally determined laminar flame speeds for primary reference fuel (PRF) mixtures of n-heptane/iso-octane and real gasoline fuel at different initial temperature and at atmospheric pressure. Nitrogen dilution effects on the laminar flame speed are also studied for selected equivalence ratios at the same conditions. A minimization of detailed kinetic model for PRF mixtures on laminar flame speed conditions was performed and the measured flame speeds were compared with numerical predictions using this model. The measured laminar flame speeds of n-decane/air mixtures at 500 K and at atmospheric pressure with and without dilution were determined. The measured flame speeds are significantly different that those predicted using existing published kinetic models, including a model validated previously against high temperature data from flow reactor, jet-stirred reactor, shock tube ignition delay, and burner stabilized flame experiments. A significant update of this model is described which

  7. Radionuclide release from spent fuel under geologic disposal conditions: An overview of experimental and theoretical work through 1985

    SciTech Connect

    Reimus, P.W.; Simonson, S.A.

    1988-04-01

    This report presents an overview of experimental and theoretical work on radionuclide release from spent fuel and uranium dioxide (UO/sub 2/) under geologic disposal conditions. The purpose of the report is to provide a source book of information that can be used to develop models that describe radionuclide release from spent fuel waste packages. Modeling activities of this nature will be conducted within the Waste Package Program (WPP) of the Department of Energy's Salt Repository Project (SRP). The topics discussed include experimental methods for investigating radionuclide release, how results have been reported from radionuclide release experiments, theoretical studies of UO/sub 2/ and actinide solubility, results of experimental studies of radionuclide release from spent fuel and UO/sub 2/ (i.e., the effects of different variables on radionuclide release), characteristics of spent fuel pertinent to radionuclide release, and status of modeling of radionuclide release from spent fuel. Appendix A presents tables of data from spent fuel radionuclide release experiments. These data have been digitized from graphs that appear in the literature. An annotated bibliography of literature on spent fuel characterization is provided in Appendix B.

  8. An experimental study of a PEM fuel cell power train for urban bus application

    NASA Astrophysics Data System (ADS)

    Corbo, P.; Migliardini, F.; Veneri, O.

    An experimental study was carried out on a fuel cell propulsion system for minibus application with the aim to investigate the main issues of energy management within the system in dynamic conditions. The fuel cell system (FCS), based on a 20 kW PEM stack, was integrated into the power train comprising DC-DC converter, Pb batteries as energy storage systems and asynchronous electric drive of 30 kW. As reference vehicle a minibus for public transportation in historical centres was adopted. A preliminary experimental analysis was conducted on the FCS connected to a resistive load through a DC-DC converter, in order to verify the stack dynamic performance varying its power acceleration from 0.5 kW s -1 to about 4 kW s -1. The experiments on the power train were conducted on a test bench able to simulate the vehicle parameters and road characteristics on specific driving cycles, in particular the European R40 cycle was adopted as reference. The "soft hybrid" configuration, which permitted the utilization of a minimum size energy storage system and implied the use of FCS mainly in dynamic operation, was compared with the "hard hybrid" solution, characterized by FCS operation at limited power in stationary conditions. Different control strategies of power flows between fuel cells, electric energy storage system and electric drive were adopted in order to verify the two above hybrid approaches during the vehicle mission, in terms of efficiencies of individual components and of the overall power train. The FCS was able to support the dynamic requirements typical of R40 cycle, but an increase of air flow rate during the fastest acceleration phases was necessary, with only a slight reduction of FCS efficiency. The FCS efficiency resulted comprised between 45 and 48%, while the overall power train efficiency reached 30% in conditions of constant stack power during the driving cycle.

  9. Experimental study on air-stream gasification of biomass micron fuel (BMF) in a cyclone gasifier.

    PubMed

    Guo, X J; Xiao, B; Zhang, X L; Luo, S Y; He, M Y

    2009-01-01

    Based on biomass micron fuel (BMF) with particle size of less than 250 microm, a cyclone gasifier concept has been considered in our laboratory for biomass gasification. The concept combines and integrates partial oxidation, fast pyrolysis, gasification, and tar cracking, as well as a shift reaction, with the purpose of producing a high quality of gas. In this paper, experiments of BMF air-stream gasification were carried out by the gasifier, with energy for BMF gasification produced by partial combustion of BMF within the gasifier using a hypostoichiometric amount of air. The effects of ER (0.22-0.37) and S/B (0.15-0.59) and biomass particle size on the performances of BMF gasification and the gasification temperature were studied. Under the experimental conditions, the temperature, gas yields, LHV of the gas fuel, carbon conversion efficiency, stream decomposition and gasification efficiency varied in the range of 586-845 degrees C, 1.42-2.21 N m(3)/kg biomass, 3806-4921 kJ/m(3), 54.44%-85.45%, 37.98%-70.72%, and 36.35%-56.55%, respectively. The experimental results showed that the gasification performance was best with ER being 3.7 and S/B being 0.31 and smaller particle, as well as H(2)-content. And the BMF gasification by air and low temperature stream in the cyclone gasifier with the energy self-sufficiency is reliable.

  10. Experimental Demonstration of Technologies for Autonomous On-Orbit Robotic Assembly

    NASA Technical Reports Server (NTRS)

    LeMaster, Edward A.; Schaechter, David B.; Carrington, Connie K.

    2006-01-01

    The Modular Reconfigurable High Energy (MRHE) program aimed to develop technologies for the automated assembly and deployment of large-scale space structures and aggregate spacecraft. Part of the project involved creation of a terrestrial robotic testbed for validation and demonstration of these technologies and for the support of future development activities. This testbed was completed in 2005, and was thereafter used to demonstrate automated rendezvous, docking, and self-assembly tasks between a group of three modular robotic spacecraft emulators. This paper discusses the rationale for the MRHE project, describes the testbed capabilities, and presents the MRHE assembly demonstration sequence.

  11. Experimental and Numerical Studies for Soot Formation in Laminar Coflow Diffusion Flames of Jet A-1 and Synthetic Jet Fuels

    NASA Astrophysics Data System (ADS)

    Saffaripour, Meghdad

    In the present doctoral thesis, fundamental experimental and numerical studies are conducted for the laminar, atmospheric pressure, sooting, coflow diffusion flames of Jet A-1 and synthetic jet fuels. The first part of this thesis presents a comparative experimental study for Jet A-1, which is a widely used petroleum-based fuel, and four synthetically produced alternative jet fuels. The main goals of this part of the thesis are to compare the soot emission levels of the alternative fuels to those of a standard fuel, Jet A-1, and to determine the effect of fuel chemical composition on soot formation characteristics. To achieve these goals, experimental measurements are constructed and performed for flame temperature, soot concentration, soot particle size, and soot aggregate structure in the flames of pre-vaporized jet fuels. The results show that a considerable reduction in soot production, compared to the standard fuel, can be obtained by using synthetic fuels which will help in addressing future regulations. A strong correlation between the aromatic content of the fuels and the soot concentration levels in the flames is observed. The second part of this thesis presents the development and experimental validation of a fully-coupled soot formation model for laminar coflow jet fuel diffusion flames. The model is coupled to a detailed kinetic mechanism to predict the chemical structure of the flames and soot precursor concentrations. This model also provides information on size and morphology of soot particles. The flames of a three-component surrogate for Jet A-1, a three-component surrogate for a synthetic jet fuel, and pure n-decane are simulated using this model. Concentrations of major gaseous species and flame temperatures are well predicted by the model. Soot volume fractions are predicted reasonably well everywhere in the flame, except near the flame centerline where soot concentrations are underpredicted by a factor of up to five. There is an excellent

  12. Parametric study of CHF data. Volume 2. A generalized subchannel CHF correlation for PWR and BWR fuel assemblies. Final report

    SciTech Connect

    Reddy, D.G.; Fighetti, C.F.

    1983-01-01

    The primary objective of this research was to develop a generalized subchannel CHF correlation based on the local fluid conditions obtained with the COBRA-IIIC thermal hydraulic subchannel code and covering PWR and BWR normal operating conditions as well as hypothetical loss-of-coolant accident (LOCA) conditions. In view of the importance of the local conditions predicted by the COBRA-IIIC code in the development of CHR correlation, the secondary objective was to improve the predictive capability of the COBRA-IIIC subchannel code. In the first phase of this study, the sensitivity of local enthalpies and local mass fluxes predicted by the COBRA-IIIC subchannel code to subcooled void correlation, bulk void correlation, two-phase friction multiplier correlation and turbulent mixing parameter was determined. In the second phase, based on the local conditions obtained with the COBRA-IIIC subchannel code, an accurate generalized subchannel CHF correlation was developed utilizing 3607 CHF data points from 65 test sections simulating PWR and BWR fuel assemblies.

  13. Conception and optimization of a membrane electrode assembly microbial fuel cell (MEA-MFC) for treatment of domestic wastewater.

    PubMed

    Lefebvre, O; Uzabiaga, A; Shen, Y J; Tan, Z; Cheng, Y P; Liu, W; Ng, H Y

    2011-01-01

    A membrane electrode assembly (MEA) for microbial fuel cells (MEA-MFC) was developed for continuous electricity production while treating domestic wastewater concurrently. It was optimized via three upgraded versions (noted α, β and γ) in terms of design (current collectors, hydrophilic separator nature) and operating conditions (hydraulic retention time, external resistance, aeration rate, recirculation). An overall rise of power by over 100% from version α to γ shows the importance of factors such as the choice of proper construction materials and prevention of short-circuits. A power of 2.5 mW was generated with a hydraulic retention time of 2.3 h when a Selemion proton exchange membrane was used as a hydrophilic separator in the MEA and 2.8 mW were attained with a reverse osmosis membrane. The MFC also showed a competitive value of internal resistance (≈40-50 Ω) as compared to the literature, especially considering its large volume (3 L). However, the operation of our system in a complete loop where the anolyte was allowed to trickle over the cathode (version γ) resulted in system failure.

  14. Simulations of Fuel Assembly and Fast-Electron Transport in Integrated Fast-Ignition Experiments on OMEGA

    NASA Astrophysics Data System (ADS)

    Solodov, A. A.; Theobald, W.; Anderson, K. S.; Shvydky, A.; Epstein, R.; Betti, R.; Myatt, J. F.; Stoeckl, C.; Jarrott, L. C.; McGuffey, C.; Qiao, B.; Beg, F. N.; Wei, M. S.; Stephens, R. B.

    2013-10-01

    Integrated fast-ignition experiments on OMEGA benefit from improved performance of the OMEGA EP laser, including higher contrast, higher energy, and a smaller focus. Recent 8-keV, Cu-Kα flash radiography of cone-in-shell implosions and cone-tip breakout measurements showed good agreement with the 2-D radiation-hydrodynamic simulations using the code DRACO. DRACO simulations show that the fuel assembly can be further improved by optimizing the compression laser pulse, evacuating air from the shell, and by adjusting the material of the cone tip. This is found to delay the cone-tip breakout by ~220 ps and increase the core areal density from ~80 mg/cm2 in the current experiments to ~500 mg/cm2 at the time of the OMEGA EP beam arrival before the cone-tip breakout. Simulations using the code LSP of fast-electron transport in the recent integrated OMEGA experiments with Cu-doped shells will be presented. Cu-doping is added to probe the transport of fast electrons via their induced Cu K-shell fluorescent emission. This material is based upon work supported by the Department of Energy National Nuclear Security Administration DE-NA0001944 and the Office of Science under DE-FC02-04ER54789.

  15. Assembly of coupled redox fuel cells using copper as electron acceptors to generate power and its in-situ retrieval

    PubMed Central

    Zhang, Hui-Min; Xu, Wei; Li, Gang; Liu, Zhan-Meng; Wu, Zu-Cheng; Li, Bo-Geng

    2016-01-01

    Energy extraction from waste has attracted much interest nowadays. Herein, a coupled redox fuel cell (CRFC) device using heavy metals, such as copper, as an electron acceptor is assembled to testify the recoveries of both electricity and the precious metal without energy consumption. In this study, a NaBH4-Cu(II) CRFC was employed as an example to retrieve copper from a dilute solution with self-electricity production. The properties of the CRFC have been characterized, and the open circuit voltage was 1.65 V with a maximum power density of 7.2 W m−2 at an initial Cu2+ concentration of 1,600 mg L−1 in the catholyte. 99.9% of the 400 mg L−1 copper was harvested after operation for 24 h, and the product formed on the cathode was identified as elemental copper. The CRFC demonstrated that useful chemicals were recovered and the electricity contained in the chemicals was produced in a self-powered retrieval process. PMID:26877144

  16. Experimental plan for the fuel-oil study. Weatherization Assistance Program: Volume 2

    SciTech Connect

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

    1992-01-01

    An up-to-date assessment of the Weatherization Assistance Program (WAP) is being performed by the US Department of Energy WAP Division and the Oak Ridge National Laboratory. Five studies form the evaluation. Major goals of the Fuel-Oil Study are to estimate the fuel oil saved by the WAP in the Northeast during the 1990 and 1991 program years, identify and quantify non-energy impacts of the WAP, assess the cost effectiveness of the WAP within this submarket, and assess factors which may cause savings and cost effectiveness to vary. The study will only analyze single-family houses in the nine states in the Northeast census region and will be carried out over two heating seasons (1990 and 1991 WAP program years). A split-winter, pre- and post-weatherization experimental design with a control group will be used. Houses will be monitored over one winter. Energy conservation measures will be installed in the weatherized houses in January of each winter by the local WAP subgrantee. One hundred twenty five weatherized houses and 75 control houses will be monitored over the 1990--1991 winter; a different set of 200 houses will be monitored over the 1991--1992 winter. The houses will be evenly distributed among 25 subgrantees. Space-heating fuel-oil consumption, indoor temperature, and outdoor temperature data will be collected for all houses. Fuel-oil delivery data will be collected for each house monitored over the 1990--1991 winter for at least a year before weatherization. The delivery data will be analyzed to determine if the accuracy of the study can be improved by collecting fuel-oil delivery data on a larger sample of houses over the 1991--1992 winter. Detailed survey information will be obtained on all the houses. This information includes descriptive details of the house and its mechanical systems, details on household size and other demographics, and occupant answers to questions regarding comfort, safety, and operation of their space-heating system and house.

  17. Direct Experimental Evaluation of the Grain Boundaries Gas Content in PWR fuels: New Insight and Perspective of the ADAGIO Technique

    SciTech Connect

    Pontillon, Y.; Noirot, J.; Caillot, L.

    2007-07-01

    Over the last decades, many analytical experiments (in-pile and out-of-pile) have underlined the active role of the inter-granular gases on the global fuel transient behavior under accidental conditions such as RIA and/or LOCA. In parallel, the improvement of fission gas release modeling in nuclear fuel performance codes needs direct experimental determination/validation regarding the local gas distribution inside the fuel sample. In this context, an experimental program, called 'ADAGIO' (French acronym for Discriminating Analysis of Accumulation of Inter-granular and Occluded Gas), has been initiated through a joint action of CEA, EDF and AREVA NP in order to develop a new device/technique for quantitative and direct measurement of local fission gas distribution within an irradiated fuel pellet. ADAGIO technique is based on the fact that fission gas inventory (intra and inter-granular parts) can be distinguished by controlled fuel oxidation, since grain boundaries oxidize faster than the bulk. The purpose of the current paper is to present both the methodology and the associated results of the ADAGIO program performed at CEA. It has been divided into two main parts: (i) feasibility (UO{sub 2} and MOX fuels), (ii) application on high burn up UO{sub 2} fuel. (authors)

  18. Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality.

    PubMed

    Rezaeian, M; Kamali, J

    2017-01-01

    Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B4C) were investigated, and the minimum required receptacle pitch of the basket was determined.

  19. Carbon-13 and proton nuclear magnetic resonance analysis of shale-derived refinery products and jet fuels and of experimental referee broadened-specification jet fuels

    NASA Technical Reports Server (NTRS)

    Dalling, D. K.; Bailey, B. K.; Pugmire, R. J.

    1984-01-01

    A proton and carbon-13 nuclear magnetic resonance (NMR) study was conducted of Ashland shale oil refinery products, experimental referee broadened-specification jet fuels, and of related isoprenoid model compounds. Supercritical fluid chromatography techniques using carbon dioxide were developed on a preparative scale, so that samples could be quantitatively separated into saturates and aromatic fractions for study by NMR. An optimized average parameter treatment was developed, and the NMR results were analyzed in terms of the resulting average parameters; formulation of model mixtures was demonstrated. Application of novel spectroscopic techniques to fuel samples was investigated.

  20. A Paradigm for the Nondestructive Assay of Spent Fuel Assemblies and Similar Large Objects, with Emphasis on the Role of Photon-Based Techniques

    NASA Astrophysics Data System (ADS)

    Bolind, Alan Michael

    2015-10-01

    The practice of nondestructive assay (NDA) of nuclear materials has, until now, been focused primarily (1) on smaller objects (2) with less fissile material and (3) with less self-generated radiation. The transition to the application of NDA to spent fuel assemblies and similar large objects violates these three conditions, thereby bringing the assumptions and paradigm of traditional NDA practice into question for the new applications. In this paper, a new paradigm for these new applications is presented which is based on the fundamental principles of nuclear engineering. It is shown that the NDA of spent fuel assemblies is mostly a three-dimensional problem that requires the integration of three independent NDA measurements in order to achieve a unique and accurate assay. The only NDA techniques that can avoid this requirement are those that analyze signals that are characteristic to specific isotopes (such as those caused by characteristic resonance interactions), and that are neither distorted nor overly attenuated by the other surrounding material. Some photon-based NDA techniques fall into this exceptional category. Such exceptional NDA techniques become essential to employ when assaying large objects that, unlike spent fuel assemblies, do not have a consistent geometry. With this new NDA paradigm, the advanced photon-based NDA techniques can be put into their proper context, and their development can thereby be properly motivated.

  1. Critical Configuration and Physics Measurements for Beryllium Reflected Assemblies of U(93.15)O₂ Fuel Rods (1.506-cm Pitch and 7-Tube Clusters)

    SciTech Connect

    Marshall, Margaret A.; Bess, John D.; Briggs, J. Blair; Murphy, Michael F.; Mihalczo, John T.

    2015-03-01

    Cadmium ratios were measured with enriched uranium metal foils at various locations in the assembly with the fuel tube at the 1.506-cm spacing. They are described in the following subsections. The experiment configuration was the same as the first critical configuration described in HEU-COMP-FAST-004 (Case 1). The experimenter placed 0.75-cm-diameter × 0.010-cm-thick 93.15%-235U-enriched uranium metal foils with and without 0.051-cm-thick cadmium covers at various locations in the core and top reflector. One part of the cadmium cover was cupshape and contained the uranium foil. The other part was a lid that fit over the exposed side of the foil when it was in the cup shaped section of the cover. As can be seen in the logbook, two runs were required to obtain all the measurements necessary for the cadmium ratio. The bare foil measurements within the top reflector were run first as part of the axial foil activation measurements. The results of this run are used for both the axial activation results and the cadmium ratios. Cadmium covered foils were then placed at the same location through the top reflector in a different run. Three pairs of bare and cadmium covered foils were also placed through the core tank. One pair was placed at the axial center of a fuel tube 11.35 cm from the center of the core. Two pairs of foils were placed on top of fuel tubes 3.02 and 12.06 cm from the center of the core. The activation of the uranium metal foils was measured after removal from the assembly using two lead shielded NaI scintillation detectors as follows. The NaI scintillators were carefully matched and had detection efficiencies for counting delayed-fission-product gamma rays with energies above 250 KeV within 5%. In all foil activation measurements, one foil at a specific location was used as a normalizing foil to remove the effects of the decay of fission products during the counting measurements with the NaI detectors. The normalization foil was placed on one Na

  2. CRITICAL CONFIGURATION AND PHYSICS MEASUREMENTS FOR BERYLLIUM REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH AND 7-TUBE CLUSTERS)

    SciTech Connect

    Margaret A. Marshall

    2014-03-01

    Cadmium ratios were measured with enriched uranium metal foils at various locations in the assembly with the fuel tube at the 1.506-cm spacing. They are described in the following subsections. The experiment configuration was the same as the first critical configuration described in HEU-COMP-FAST-004 (Case 1). The experimenter placed 0.75-cm-diameter × 0.010-cm-thick 93.15%-235U-enriched uranium metal foils with and without 0.051-cm-thick cadmium covers at various locations in the core and top reflector. One part of the cadmium cover was cupshape and contained the uranium foil. The other part was a lid that fit over the exposed side of the foil when it was in the cup shaped section of the cover. As can be seen in the logbook, two runs were required to obtain all the measurements necessary for the cadmium ratio. The bare foil measurements within the top reflector were run first as part of the axial foil activation measurements. The results of this run are used for both the axial activation results and the cadmium ratios. Cadmium covered foils were then placed at the same location through the top reflector in a different run. Three pairs of bare and cadmium covered foils were also placed through the core tank. One pair was placed at the axial center of a fuel tube 11.35 cm from the center of the core. Two pairs of foils were placed on top of fuel tubes 3.02 and 12.06 cm from the center of the core. The activation of the uranium metal foils was measured after removal from the assembly using two lead shielded NaI scintillation detectors as follows. The NaI scintillators were carefully matched and had detection efficiencies for counting delayed-fission-product gamma rays with energies above 250 KeV within 5%. In all foil activation measurements, one foil at a specific location was used as a normalizing foil to remove the effects of the decay of fission products during the counting measurements with the NaI detectors. The normalization foil was placed on one Na

  3. CRITICAL CONFIGURATION AND PHYSICS MEASUREMENTS FOR BERYLLIUM REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH)

    SciTech Connect

    Margaret A. Marshall

    2013-03-01

    Cadmium ratios were measured with enriched uranium metal foils at various locations in the assembly with the fuel tube at the 1.506-cm spacing. They are described in the following subsections. The experiment configuration was the same as the first critical configuration described in HEU-COMP-FAST-004 (Case 1). The experimenter placed 0.75-cm-diameter × 0.010-cm-thick 93.15%-235U-enriched uranium metal foils with and without 0.051-cm-thick cadmium covers at various locations in the core and top reflector. One part of the cadmium cover was cupshape and contained the uranium foil. The other part was a lid that fit over the exposed side of the foil when it was in the cup shaped section of the cover. As can be seen in the logbook, two runs were required to obtain all the measurements necessary for the cadmium ratio. The bare foil measurements within the top reflector were run first as part of the axial foil activation measurements. The results of this run are used for both the axial activation results and the cadmium ratios. Cadmium covered foils were then placed at the same location through the top reflector in a different run. Three pairs of bare and cadmium covered foils were also placed through the core tank. One pair was placed at the axial center of a fuel tube 11.35 cm from the center of the core. Two pairs of foils were placed on top of fuel tubes 3.02 and 12.06 cm from the center of the core. The activation of the uranium metal foils was measured after removal from the assembly using two lead shielded NaI scintillation detectors as follows. The NaI scintillators were carefully matched and had detection efficiencies for counting delayed-fission-product gamma rays with energies above 250 KeV within 5%. In all foil activation measurements, one foil at a specific location was used as a normalizing foil to remove the effects of the decay of fission products during the counting measurements with the NaI detectors. The normalization foil was placed on one Na

  4. Modeling and experimental studies to optimize the performance of a hydrogen - bromine fuel cell

    NASA Astrophysics Data System (ADS)

    Yarlagadda, Venkata Raviteja

    The regenerative Hydrogen-Bromine (H2-Br 2) fuel cells are considered to be one of the viable systems for large scale energy storage because of their high energy conversion efficiency, flexible operation, highly reversible reactions and low capital cost. The preliminary performance of a H2-Br2 fuel cell using both conventional as well as novel materials (Nafion and electrospun composite membranes along with platinum and rhodium sulfide electrocatalysts) was discussed. A maximum power density of 0.65 W/cm2 was obtained with a thicker Br 2 electrode (780 mum) and cell temperature of 45°C. The active area and wetting characteristics of Br2 electrodes were improved upon by either pre-treating with HBr or boiling them in de-ionized water. On the other hand, similar or better performances were obtained using dual fiber electrospun composite membranes (maximum power densities of 0.61 W/cm2 and 0.45 W/cm2 obtained with 25 mum and 65 mum electrospun membranes at 45°C) versus using Nafion membranes (maximum power densities of 0.52 W/cm 2 and 0.41 W/cm2 obtained with Nafion 212 and Nafion 115 membranes at 45°C). The rhodium sulfide (RhxSy) electrocatalyst proved to be more stable in the presence of HBr/Br2 than pure Pt. However, the H2 oxidation activity on RhxS y was quite low compared to that of Pt. In conclusion, a stable H 2 electrocatalyst that can match the hydrogen oxidation activity obtained with Pt and a membrane with low Br2/Br- permeability are essential to prolong the lifetime of a H2-Br2 fuel cell. A 1D mathematical model was developed to serve as a theoretical guiding tool for the experimental studies. The impact of convective and diffusive transport and kinetic rate on the performance of a H2-Br2 fuel cell is shown in this study. Of the two flow designs (flow-by and flow-through) incorporated in this study, the flow-through design demonstrated better performance, which can be attributed to the dominant convective transport inside the porous electrode. Both

  5. An Experimental and Chemical Kinetics Study of the Combustion of Syngas and High Hydrogen Content Fuels

    SciTech Connect

    Santoro, Robers; Dryer, Frederick; Ju, Yiguang

    2013-09-30

    An integrated and collaborative effort involving experiments and complementary chemical kinetic modeling investigated the effects of significant concentrations of water and CO2 and minor contaminant species (methane [CH4], ethane [C2H6], NOX, etc.) on the ignition and combustion of HHC fuels. The research effort specifically addressed broadening the experimental data base for ignition delay, burning rate, and oxidation kinetics at high pressures, and further refinement of chemical kinetic models so as to develop compositional specifications related to the above major and minor species. The foundation for the chemical kinetic modeling was the well validated mechanism for hydrogen and carbon monoxide developed over the last 25 years by Professor Frederick Dryer and his co-workers at Princeton University. This research furthered advance the understanding needed to develop practical guidelines for realistic composition limits and operating characteristics for HHC fuels. A suite of experiments was utilized that that involved a high-pressure laminar flow reactor, a pressure-release type high-pressure combustion chamber and a high-pressure turbulent flow reactor.

  6. Evaluation of the potential of various aquatic eco-systems in harnessing bioelectricity through benthic fuel cell: effect of electrode assembly and water characteristics.

    PubMed

    Venkata Mohan, S; Srikanth, S; Veer Raghuvulu, S; Mohanakrishna, G; Kiran Kumar, A; Sarma, P N

    2009-04-01

    Six different types of ecological water bodies were evaluated to assess their potential to generate bioelectricity using benthic type fuel cell assemblies. Experiments were designed with various combinations of electrode assemblies, surface area of anode and anodic materials. Among the 32 experiments conducted, nine combinations evidenced stable electron-discharge/current. Nature, flow conditions and characteristics of water bodies showed significant influence on the power generation apart from electrode assemblies, surface area of anode and anodic material. Stagnant water bodies showed comparatively higher power output than the running water bodies. Placement of cathode on algal mat (as bio-cathode) documented several folds increment in power output. Electron-discharge started at 1000 Omega resistance in polluted water bodies (Nacaharam cheruvu, Hussain Sagar lake Musi river), whereas, in relatively less polluted water bodies (Uppal pond/stream, Godavari river) electron-discharge was observed at low resistances (500/750 Omega).

  7. A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

    DOE PAGES

    Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.; ...

    2015-12-01

    In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

  8. A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

    SciTech Connect

    Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.; Grape, Sophie; Hendricks, John S.; Henzl, Vladimir; Swinhoe, Martyn T.

    2015-12-01

    In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is be ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.

  9. Experimental investigation of the ignition and flammability limits of various hydrocarbon fuels in a two-dimensional solid-fuel ramjet. Master's thesis

    SciTech Connect

    Wooldridge, R.C.

    1987-06-01

    An experimental investigation was conducted to study the effects of inlet step height on ignition and flammability limits and recirculation-zone and boundary-layer combustion phenomena of various hydrocarbon fuels. A windowed two-dimensional solid-fuel ramjet (SFRJ) was utilized. Hydrocarbon fuels were burned under conditions similar to the actual flight were studied using a variable geometry inlet, an automatic data acquisition system, and high speed motion pictures of the interior of the combustion chamber at the recirculation zone and the boundary layer development region. Data was obtained at a mass flux of 0.2 1bm/in2-sec at a nominal air inlet temperature of 1000 R with pressures ranging from 100 to 150 psia.

  10. Design fabrication and nondestructive testing of six experimental AGCarb/Intermold 3 cylinder assemblies

    NASA Technical Reports Server (NTRS)

    Thacher, E. F.

    1972-01-01

    Six subscale Intermold cylinder assemblies with a total of twelve different concepts for transition to AGCarb were fabricated. Three of the cylinder assemblies were made by helically winding the hoop fibers and three were of orthogonal configuration. The fabrication process is summarized and details of each manufacturing method are given. The objectives of the test program were to: (1) demonstrate the fabricability of the Intermold 3 subscale flanges, (2) produce an integral transition from Intermold 3 to AGCarb material, (3) define a workable manufacturing process, and (4) identify a best suited inspection method. The objectives were met and the results are described.

  11. Robotic Manufacturing of 5.5 Meter Cryogenic Fuel Tank Dome Assemblies for the NASA Ares I Rocket

    NASA Technical Reports Server (NTRS)

    Jones, Ronald E.

    2012-01-01

    The Ares I rocket is the first launch vehicle scheduled for manufacture under the National Aeronautic and Space Administration's (NASA's) Constellation program. A series of full-scale Ares I development articles have been constructed on the Robotic Weld Tool at the NASA George C. Marshall Space Flight Center in Huntsville, Alabama. The Robotic Weld Tool is a 100 ton, 7-axis, robotic manufacturing system capable of machining and friction stir welding large-scale space hardware. This presentation will focus on the friction stir welding of 5.5m diameter cryogenic fuel tank components; specifically, the liquid hydrogen forward dome (LH2 MDA), the common bulkhead manufacturing development articles (CBMDA) and the thermal protection system demonstration dome (TPS Dome). The LH2 MDA was the first full-scale, flight-like Ares I hardware produced under the Constellation Program. It is a 5.5m diameter elliptical dome assembly consisting of eight gore panels, a y-ring stiffener and a manhole fitting. All components are made from aluminumlithium alloy 2195. Conventional and self-reacting friction stir welding was used on this article. An overview of the manufacturing processes will be discussed. The LH2 MDA is the first known fully friction stir welded dome ever produced. The completion of four Common Bulkhead Manufacturing Development Articles (CBMDA) and the TPS Dome will also be highlighted. Each CBMDA and the TPS Dome consists of a 5.5m diameter spun-formed dome friction stir welded to a y-ring stiffener. The domes and y-rings are made of aluminum 2014 and 2219 respectively. The TPS Dome has an additional aluminum alloy 2195 barrel section welded to the y-ring. Manufacturing solutions will be discussed including "fixtureless" welding with self reacting friction stir welding.

  12. A beam-modification assembly for experimental neutron capture therapy of brain tumors

    SciTech Connect

    Slatkin, D.N.; Kalef-Ezra, J.A.; Saraf, S.K.; Joel, D.D.

    1989-01-01

    Recent attempts to treat intracerebral rat gliomas by boron neutron capture therapy (BNCT) have been somewhat disappointing, perhaps in part because of excessive whole-body and nasopharyngeal irradiation. Intracerebral rat gliomas were treated by BNCT with more success using a new beam-modification assembly. 3 refs., 2 figs.

  13. Performance of Four Experimental High-btu-per-gallon Fuels in a Single Turbojet Combustor

    NASA Technical Reports Server (NTRS)

    Jonash, Edmund R; Metzler, Allen; Butze, Helmut F

    1955-01-01

    Performance characteristics of four hydrocarbon fuels having high Btu per gallon were determined in a single turbojet combustor. At simulated low-altitude operating conditions, the fuels with high Btu per gallon generally produced more carbon than did JP-4 and JP-5 fuels. The deposits were reduced appreciably with a fuel-oil additive. At high-altitude conditions, the high Btu-per-gallon fuels gave lower efficiencies than did JP-4 or JP-5 fuels. No attempts were made to improve performance by combustor design modification.

  14. Pyroprocessing of Oxidized Sodium-Bonded Fast Reactor Fuel -- an Experimental Study of Treatment Options for Degraded EBR-II Fuel

    SciTech Connect

    S. D. Herrmann; L. A. Wurth; N. J. Gese

    2013-09-01

    An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electrometallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li2O at 650 °C with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. The experimental study illustrated how zirconium oxide and sodium oxide present different challenges to a lithium-based electrolytic reduction system for conversion of select metal oxides to metal.

  15. Syngas suitability for solid oxide fuel cells applications produced via biomass steam gasification process: Experimental and modeling analysis

    NASA Astrophysics Data System (ADS)

    Pieratti, Elisa; Baratieri, Marco; Ceschini, Sergio; Tognana, Lorenzo; Baggio, Paolo

    The technologies and the processes for the use of biomass as an energy source are not always environmental friendly. It is worth to develop approaches aimed at a more sustainable exploitation of biomass, avoiding whenever possible direct combustion and rather pursuing fuel upgrade paths, also considering direct conversion to electricity through fuel cells. In this context, it is of particular interest the development of the biomass gasification technology for synthesis gas (i.e., syngas) production, and the utilization of the obtained gas in fuel cells systems, in order to generate energy from renewable resources. Among the different kind of fuel cells, SOFCs (solid oxide fuel cells), which can be fed with different type of fuels, seem to be also suitable for this type of gaseous fuel. In this work, the syngas composition produced by means of a continuous biomass steam gasifier (fixed bed) has been characterized. The hydrogen concentration in the syngas is around 60%. The system is equipped with a catalytic filter for syngas purification and some preliminary tests coupling the system with a SOFCs stack are shown. The data on the syngas composition and temperature profile measured during the experimental activity have been used to calibrate a 2-dimensional thermodynamic equilibrium model.

  16. Radiological characteristics of light-water reactor spent fuel: A literature survey of experimental data. [82 references

    SciTech Connect

    Roddy, J.W.; Mailen, J.C.

    1987-12-01

    This survey brings together the experimentally determined light-water reactor spent fuel data comprising radionuclide composition, decay heat, and photon and neutron generation rates as identified in a literature survey. Many citations compare these data with values calculated using a radionuclide generation and depletion computer code, ORIGEN, and these comparisons have been included. ORIGEN is a widely recognized method for estimating the actinide, fission product, and activation product contents of irradiated reactor fuel, as well as the resulting heat generation and radiation levels. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of spent fuel. 82 refs., 4 figs., 17 tabs.

  17. Combined Theoretical and Experimental Analysis of Processes Determining Cathode Performance in Solid Oxide Fuel Cells

    SciTech Connect

    Kukla, Maija M.; Kotomin, Eugene Alexej; Merkle, R.; Mastrikov, Yuri; Maier, J.

    2013-02-11

    Solid oxide fuel cells (SOFC) are under intensive investigation since the 1980’s as these devices open the way for ecologically clean direct conversion of the chemical energy into electricity, avoiding the efficiency limitation by Carnot’s cycle for thermochemical conversion. However, the practical development of SOFC faces a number of unresolved fundamental problems, in particular concerning the kinetics of the electrode reactions, especially oxygen reduction reaction. We review recent experimental and theoretical achievements in the current understanding of the cathode performance by exploring and comparing mostly three materials: (La,Sr)MnO3 (LSM), (La,Sr)(Co,Fe)O3 (LSCF) and (Ba,Sr)(Co,Fe)O3 (BSCF). Special attention is paid to a critical evaluation of advantages and disadvantages of BSCF, which shows the best cathode kinetics known so far for oxides. We demonstrate that it is the combined experimental and theoretical analysis of all major elementary steps of the oxygen reduction reaction which allows us to predict the rate determining steps for a given material under specific operational conditions and thus control and improve SOFC performance.

  18. Combined theoretical and experimental analysis of processes determining cathode performance in solid oxide fuel cells.

    PubMed

    Kuklja, M M; Kotomin, E A; Merkle, R; Mastrikov, Yu A; Maier, J

    2013-04-21

    Solid oxide fuel cells (SOFC) are under intensive investigation since the 1980's as these devices open the way for ecologically clean direct conversion of the chemical energy into electricity, avoiding the efficiency limitation by Carnot's cycle for thermochemical conversion. However, the practical development of SOFC faces a number of unresolved fundamental problems, in particular concerning the kinetics of the electrode reactions, especially oxygen reduction reaction. We review recent experimental and theoretical achievements in the current understanding of the cathode performance by exploring and comparing mostly three materials: (La,Sr)MnO3 (LSM), (La,Sr)(Co,Fe)O3 (LSCF) and (Ba,Sr)(Co,Fe)O3 (BSCF). Special attention is paid to a critical evaluation of advantages and disadvantages of BSCF, which shows the best cathode kinetics known so far for oxides. We demonstrate that it is the combined experimental and theoretical analysis of all major elementary steps of the oxygen reduction reaction which allows us to predict the rate determining steps for a given material under specific operational conditions and thus control and improve SOFC performance.

  19. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. B. Grover

    2007-05-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing.1,2 The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  20. Final Report for NFE-07-00912: Development of Model Fuels Experimental Engine Data Base & Kinetic Modeling Parameter Sets

    SciTech Connect

    Bunting, Bruce G

    2012-10-01

    The automotive and engine industries are in a period of very rapid change being driven by new emission standards, new types of after treatment, new combustion strategies, the introduction of new fuels, and drive for increased fuel economy and efficiency. The rapid pace of these changes has put more pressure on the need for modeling of engine combustion and performance, in order to shorten product design and introduction cycles. New combustion strategies include homogeneous charge compression ignition (HCCI), partial-premixed combustion compression ignition (PCCI), and dilute low temperature combustion which are being developed for lower emissions and improved fuel economy. New fuels include bio-fuels such as ethanol or bio-diesel, drop-in bio-derived fuels and those derived from new crude oil sources such as gas-to-liquids, coal-to-liquids, oil sands, oil shale, and wet natural gas. Kinetic modeling of the combustion process for these new combustion regimes and fuels is necessary in order to allow modeling and performance assessment for engine design purposes. In this research covered by this CRADA, ORNL developed and supplied experimental data related to engine performance with new fuels and new combustion strategies along with interpretation and analysis of such data and consulting to Reaction Design, Inc. (RD). RD performed additional analysis of this data in order to extract important parameters and to confirm engine and kinetic models. The data generated was generally published to make it available to the engine and automotive design communities and also to the Reaction Design Model Fuels Consortium (MFC).

  1. Experimental research on the rotating detonation in gaseous fuels-oxygen mixtures

    NASA Astrophysics Data System (ADS)

    Kindracki, J.; Wolański, P.; Gut, Z.

    2011-04-01

    An experimental study on rotating detonation is presented in this paper. The study was focused on the possibility of using rotating detonation in a rocket engine. The research was divided into two parts: the first part was devoted to obtaining the initiation of rotating detonation in fuel-oxygen mixture; the second was aimed at determination of the range of propagation stability as a function of chamber pressure, composition, and geometry. Additionally, thrust and specific impulse were determined in the latter stage. In the paper, only rich mixture is described, because using such a composition in rocket combustion chambers maximizes the specific impulse and thrust. In the experiments, two kinds of geometry were examined: cylindrical and cylindrical-conic, the latter can be simulated by a simple aerospike nozzle. Methane, ethane, and propane were used as fuel. The pressure-time courses in the manifolds and in the chamber are presented. The thrust-time profile and detonation velocity calculated from measured pressure peaks are shown. To confirm the performance of a rocket engine with rotating detonation as a high energy gas generator, a model of a simple engine was designed, built, and tested. In the tests, the model of the engine was connected to the dump tank. This solution enables different environmental conditions from a range of flight from 16 km altitude to sea level to be simulated. The obtained specific impulse for pressure in the chamber of max. 1.2 bar and a small nozzle expansion ratio of about 3.5 was close to 1,500 m/s.

  2. Experimental comparative study of doublet and triplet impinging atomization of gelled fuel based on PIV

    NASA Astrophysics Data System (ADS)

    Yang, Jian-lu; Li, Ning; Weng, Chun-sheng

    2016-10-01

    Gelled propellant is promising for future aerospace application because of its combination of the advantages of solid propellants and liquid propellants. An effort was made to reveal the atomization properties of gelled fuel by particle image velocimetry (PIV) system. The gelled fuel which was formed by gasoline and Nano-silica was atomized using a like-doublet impingement injector and an axisymmetric like-triplet impingement injector. The orifice diameter and length of the nozzle used in this work were of 0.8mm, 4.8mm, respectively. In the impinging spray process, the impingement angles were set at 90° and 120°, and the injection pressures were of 0.50MPa and 1.00MPa. The distance from the exit of the orifice to the impingement point was fixed at 9.6mm. In this study, high-speed visualization and temporal resolution particle image velocimetry techniques were employed to investigate the impingement atomization characteristics. The experimental investigation demonstrated that a long narrow high speed droplets belt formed around the axis of symmetry in the like-doublet impinging atomization area. However, there was no obvious high-speed belt with impingement angle 2θ = 90° and two high-speed belts appeared with impingement angle 2θ = 120° in the like-doublet impingement spray field. The high droplet velocity zone of the like-doublet impingement atomization symmetrically distributed around the central axis, and that of the like-triplet impingement spray deflected to the left of the central axis - opposite of injector. Although the droplets velocity distribution was asymmetry of like-triplet impingement atomization, the injectors were arranged like axisymmetric conical shape, and the cross section of spray area was similar to a circle rather than a narrow rectangle like the like-doublet impingement atomization.

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

    NASA Astrophysics Data System (ADS)

    Forquin, P.

    2010-06-01

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

  4. Experimental Studies on the Self-Shielding Effect in Fissile Fuel Breeding Measurement in Thorium Oxide Pellets Irradiated with 14 MeV Neutrons

    NASA Astrophysics Data System (ADS)

    Mitul, Abhangi; Nupur, Jain; Rajnikant, Makwana; Sudhirsinh, Vala; Shrichand, Jakhar; K. Basu, T.; V. S. Rao, C.

    2013-02-01

    The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 fissile fuel from fertile material Thorium-232. In order to estimate the amount of U-233 produced, experiments are carried out by irradiating thorium dioxide pellets with neutrons produced from a 14 MeV neutron generator. The objective of the present work is to measure the reaction rates of 232Th + 1n → 233Th → 233Pa → 233U in different pellet thicknesses to study the self-shielding effects and adopt a procedure for correction. An appropriate assembly consisting of high-density polyethylene is designed and fabricated to slow down the high-energy neutrons, in which Thorium pellets are irradiated. The amount of fissile fuel (233U) produced is estimated by measuring the 312 keV gammas emitted by Protactinium-233 (half-life of 27 days). A calibrated High Purity Germanium (HPGe) detector is used to measure the gamma ray spectrum. The amount of 233U produced by Th232 (n, γ) is calculated using MCNP code. The self-shielding effect is evaluated by calculating the reaction rates for different foil thickness. MCNP calculation results are compared with the experimental values and appropriate correction factors are estimated for self-shielding of neutrons and absorption of gamma rays.

  5. Identification to a breached fuel pin in the IEM cell

    SciTech Connect

    McGuinness, P.W.; Kalk, J.J.; Hicks, D.F.

    1987-01-01

    Novel methods were successfully employed to identify one breached fuel pin in a 217-pin fuel assembly. The assembly was an experiment that had been irradiated at the Fast Flux Test Facility (FFTF), an experimental liquid-metal reactor operated by Westinghouse Hanford Company for the US Dept. of Energy. A fuel assembly known to contain breached fuel pins was removed from the sodium-cooled FFTF reactor in November 1984. Later, this assembly was brought into the FFTF's Interim Examination and Maintenance (IEM) cell to be disassembled and, for the first time ever at FFTF, to identify a breached fuel pin. The synergistic evaluation of the four different verification techniques - visual examination, cladding swipe activity, wash water radiochemistry, and pin weight - provided rapid and positive identification. The capability to perform future detective work of this kind has been conclusively demonstrated.

  6. End-to-end self-assembly of gold nanorods in isopropanol solution: experimental and theoretical studies

    NASA Astrophysics Data System (ADS)

    Gordel, M.; Piela, K.; Kołkowski, R.; Koźlecki, T.; Buckle, M.; Samoć, M.

    2015-12-01

    We describe here a modification of properties of colloidal gold nanorods (NRs) resulting from the chemical treatment used to carry out their transfer into isopropanol (IPA) solution. The NRs acquire a tendency to attach one to another by their ends (end-to-end assembly). We focus on the investigation of the change in position and shape of the longitudinal surface plasmon (l-SPR) band after self-assembly. The experimental results are supported by a theoretical calculation, which rationalizes the dramatic change in optical properties when the NRs are positioned end-to-end at short distances. The detailed spectroscopic characterization performed at the consecutive stages of transfer of the NRs from water into IPA solution revealed the features of the interaction between the polymers used as ligands and their contribution to the final stage, when the NRs were dispersed in IPA solution. The efficient method of aligning the NRs detailed here may facilitate applications of the self-assembled NRs as building blocks for optical materials and biological sensing.

  7. Concept report: Experimental vector magnetograph (EXVM) operational configuration balloon flight assembly

    NASA Technical Reports Server (NTRS)

    1993-01-01

    The observational limitations of earth bound solar studies has prompted a great deal of interest in recent months in being able to gain new scientific perspectives through, what should prove to be, relatively low cost flight of the magnetograph system. The ground work done by TBE for the solar balloon missions (originally planned for SOUP and GRID) as well as the rather advanced state of assembly of the EXVM has allowed the quick formulation of a mission concept for the 30 cm system currently being assembled. The flight system operational configuration will be discussed as it is proposed for short duration flight (on the order of one day) over the continental United States. Balloon hardware design requirements used in formulation of the concept are those set by the National Science Balloon Facility (NSBF), the support agency under NASA contract for flight services. The concept assumes that the flight hardware assembly would come together from three development sources: the scientific investigator package, the integration contractor package, and the NSBF support system. The majority of these three separate packages can be independently developed; however, the computer control interfaces and telemetry links would require extensive preplanning and coordination. A special section of this study deals with definition of a dedicated telemetry link to be provided by the integration contractor for video image data for pointing system performance verification. In this study the approach has been to capitalize to the maximum extent possible on existing hardware and system design. This is the most prudent step that can be taken to reduce eventual program cost for long duration flights. By fielding the existing EXVM as quickly as possible, experience could be gained from several short duration flight tests before it became necessary to commit to major upgrades for long duration flights of this system or of the larger 60 cm version being considered for eventual development.

  8. A quantitative model and the experimental evaluation of the liquid fuel layer for the downward flame spread of XPS foam.

    PubMed

    Luo, Shengfeng; Xie, Qiyuan; Tang, Xinyi; Qiu, Rong; Yang, Yun

    2017-05-05

    The objective of this work is to investigate the distinctive mechanisms of downward flame spread for XPS foam. It was physically considered as a moving down of narrow pool fire instead of downward surface flame spread for normal solids. A method was developed to quantitatively analyze the accumulated liquid fuel based on the experimental measurement of locations of flame tips and burning rates. The results surprisingly showed that about 80% of the generated hot liquid fuel remained in the pool fire during a certain period. Most of the consumed solid XPS foam didn't really burn away but transformed as the liquid fuel in the downward moving pool fire, which might be an important promotion for the fast fire development. The results also indicated that the dripping propensity of the hot liquid fuel depends on the total amount of the hot liquid accumulated in the pool fire. The leading point of the flame front curve might be the breach of the accumulated hot liquid fuel if it is enough for dripping. Finally, it is suggested that horizontal noncombustible barriers for preventing the accumulation and dripping of liquid fuel are helpful for vertical confining of XPS fire.

  9. Experimental investigation and numerical comparison of the performance of a proton exchange membrane fuel cell at different channel geometry

    NASA Astrophysics Data System (ADS)

    Khazaee, I.

    2015-08-01

    In this study, the performance of a PEM fuel cell is investigated experimentally and numerically by changing the geometry of the channels. At first an experimental setup is used and three different fuel cells with rectangular, elliptical and triangular serpentine channels are constructed. The active area of each cell is 25 cm2 that its weight is 1,300 g. The material of the gas diffusion layer is carbon clothes, the membrane is nafion 117 and the catalyst layer is a plane with 0.004 g cm-2 platinum. Then a complete three-dimensional model for fuel cell is used to investigate the effect of using this channels geometry on the performance. The proposed model is a full cell model, which includes all the parts of the PEM fuel cell, flow channels, gas diffusion electrodes, catalyst layers and the membrane. Coupled transport and electrochemical kinetics equations are solved in a single domain; therefore no interfacial boundary condition is required at the internal boundaries between cell components. The results show that the predicted polarization curves by using this model are in good agreement with the experimental results. Also the results show that when the geometry of channel is rectangular the performance of the cell is better than the triangular and elliptical channel.

  10. A technical review of non-destructive assay research for the characterization of spent nuclear fuel assemblies being conducted under the US DOE NGSI

    SciTech Connect

    Croft, Stephen; Tobin, Stephen J

    2010-12-06

    There is a growing belief that expansion of nuclear energy generation will be needed in the coming decades as part of a mixed supply chain to meet global energy demand. At stake is the health of the economic engine that delivers human prosperity. As a consequence renewed interest is being paid to the safe management of spent nuclear fuel (SNF) and the plutonium it contains. In addition to being an economically valuable resource because it can be used to construct explosive devices, Pu must be placed on an inventory and handled securely. A multiinstitutional team of diverse specialists has been assembled under a project funded by the US Department of Energy (DOE) Next Generation Safeguards Initiative (NGSI) to address ways to nondestructively quantify the plutonium content of spent nuclear fuel assemblies, and to also detect the potential diversion of pins from those assemblies. Studies are underway using mostly Monte Carlo tools to assess the feasibility, individual and collective performance capability of some fourteen nondestructive assay methods. Some of the methods are familiar but are being applied in a new way against a challenging target which is being represented with a higher degree of realism in simulation space than has been done before, while other methods are novel. In this work we provide a brief review of the techniques being studied and highlight the main achievements to date. We also draw attention to the deficiencies identified in for example modeling capability and available basic nuclear data. We conclude that this is an exciting time to be working in the NDA field and that much work, both fundamental and applied, remains ahead if we are to advance the state of the practice to meet the challenges posed to domestic and international safeguards by the expansion of nuclear energy together with the emergence of alternative fuel cycles.

  11. Coral recruitment onto an experimental pulverised fuel ash-concrete artificial reef.

    PubMed

    Lam, Katherine K Y

    2003-05-01

    An experimental artificial reef was deployed in December 1993 at Hoi Ha Wan Marine Park, Hong Kong. This is the first study documenting natural scleractinian coral recruitment onto a stabilised pulverised fuel ash (PFA)-concrete artificial reef. Visible recruits were first recorded 9-10 months after the placement of reef blocks, i.e., in the autumn of 1994. Two scleractinians, Oulastrea crispata and Culicia japonica, were recruited. The recruit density of the former was much greater than the latter. The spatial recruitment pattern of the corals was observed to be affected by the orientation of the attaching surface. O. crispata settled predominantly on the undersides of the reef blocks. There was an edge effect on O. crispata recruitment. C. japonica, however, had a preference for exposed surfaces. O. crispata did not show a preference for block composition whereas C. japonica favoured blocks with high (75% by volume) PFA levels. This shows that PFA-concrete is a potential substratum for artificial reef construction, especially when such reefs aim at rehabilitating corals.

  12. Experimental study of the oxidation of large surrogates for diesel and biodiesel fuels

    SciTech Connect

    Hakka, Mohammed Hichem; Glaude, Pierre-Alexandre; Herbinet, Olivier; Battin-Leclerc, Frederique

    2009-11-15

    The experimental study of the oxidation of two blend surrogates for diesel and biodiesel fuels, n-decane/n-hexadecane and n-decane/methyl palmitate (74/26 mol/mol), has been performed in a jet-stirred reactor over a wide range of temperatures covering both low, and high-temperature regions (550-1100 K), at a residence time of 1.5 s, at quasi atmospheric pressure with high dilution in helium (hydrocarbon inlet mole fraction of 0.002) and at stoichiometric conditions. Numerous reaction products have been identified and quantified. At low and intermediate temperatures (less than 1000 K), the formation of oxygenated species such as cyclic ethers, aldehydes and ketones has been observed for n-decane, n-hexadecane, and methyl palmitate. At higher temperature, the formation of these species was not observed any more, and small amounts of unsaturated species (olefins and unsaturated methyl esters) have been detected. Results obtained with methyl palmitate and n-hexadecane have been compared in order to highlight similarities and differences between large n-alkanes and methyl esters. (author)

  13. Experimental dissection of oxygen transport resistance in the components of a polymer electrolyte membrane fuel cell

    NASA Astrophysics Data System (ADS)

    Oh, Hwanyeong; Lee, Yoo il; Lee, Guesang; Min, Kyoungdoug; Yi, Jung S.

    2017-03-01

    Oxygen transport resistance is a major obstacle for obtaining high performance in a polymer electrolyte membrane fuel cell (PEMFC). To distinguish the major components that inhibit oxygen transport, an experimental method is established to dissect the oxygen transport resistance of the components of the PEMFC, such as the substrate, micro-porous layer (MPL), catalyst layer, and ionomer film. The Knudsen numbers are calculated to determine the types of diffusion mechanisms at each layer by measuring the pore sizes with either mercury porosimetry or BET analysis. At the under-saturated condition where condensation is mostly absent, the molecular diffusion resistance is dissected by changing the type of inert gas, and ionomer film permeation is separated by varying the inlet gas humidity. Moreover, the presence of the MPL and the variability of the substrate thickness allow the oxygen transport resistance at each component of a PEMFC to be dissected. At a low relative humidity of 50% and lower, an ionomer film had the largest resistance, while the contribution of the MPL was largest for the other humidification conditions.

  14. In vitro toxicities of experimental jet fuel system ice-inhibiting agents.

    PubMed

    Geiss, K T; Frazier, J M

    2001-07-02

    One research emphasis within the Department of Defense has been to seek the replacement of operational compounds with alternatives that pose less potential risk to human and ecological systems. Alternatives to glycol ethers, such as diethylene glycol monomethyl ether (M-DE), were investigated for use as jet fuel system ice-inhibiting agents (FSIIs). This group of chemicals includes three derivatives of 1,3-dioxolane-4-methanol (M-1, M-2, and M-3) and a 1,3-dioxane (M-27). In addition, M-DE was evaluated as a reference compound. Our approach was to implement an in vitro test battery based on primary rat hepatocyte cultures to perform initial toxicity evaluations. Hepatocytes were exposed to experimental chemicals (0, 0.001, 0.01, 0.1, 1, 10 mM dosages) for periods up to 24 h. Samples were assayed for lactate dehydrogenase (LDH) release, MTT dye reduction activity, glutathione level, and rate of protein synthesis as indicators of toxicity. Of the compounds tested, M-1, especially at the 10-mM dose, appeared to be more potent than the other chemicals, as measured by these toxicity assays. M-DE, the current FSII, elicited little response in the toxicity assays. Although some variations in toxicity were observed at the 10-mM dose, the in vitro toxicities of the chemicals tested (except for M-1) were not considerably greater than that of M-DE.

  15. Preparation of a self-assembled organosilane coating on carbon black as a catalyst support in polymer electrolyte membrane fuel cells

    NASA Astrophysics Data System (ADS)

    Lee, Woong Hee; Seo, Jungmok; Lee, Taeyoon; Kim, Hansung

    2015-01-01

    A novel method is developed to increase the resistance to electrochemical carbon corrosion using a self-assembled organosilane coating of dodecyltrichlorosilane (DTS) on carbon black (CB). This process successfully creates a hydrophobic coating on the hydrophilic surface of carbon black without poisoning Pt nanoparticles. The hydrophobic DTS coating improves the performance of CB in fuel cells by enhancing the mass transfer rate. Following carbon corrosion tests, on-line mass spectrometry shows that this DTS coating improves the electrochemical carbon corrosion resistance of CB by increasing the contact resistance of water, which is necessary for electrochemical carbon corrosion. Thus, this DTS coating is a very effective means to improve the performance and durability of fuel cells.

  16. Investigations of a Combustor Using a 9-Point Swirl-Venturi Fuel Injector: Recent Experimental Results

    NASA Technical Reports Server (NTRS)

    Hicks, Yolanda R.; Heath, Christopher M.; Anderson, Robert C.; Tacina, Kathleen M.

    2012-01-01

    This paper explores recent results obtained during testing in an optically-accessible, JP8-fueled, flame tube combustor using baseline Lean Direct Injection (LDI) research hardware. The baseline LDI geometry has nine fuel/air mixers arranged in a 3 x 3 array. Results from this nine-element array include images of fuel and OH speciation via Planar Laser-Induced Fluorescence (PLIF), which describe fuel spray pattern and reaction zones. Preliminary combustion temperatures derived from Stokes/Anti-Stokes Spontaneous Raman Spectroscopy are also presented. Other results using chemiluminescence from major combustion radicals such as CH* and C2* serve to identify the primary reaction zone, while OH PLIF shows the extent of reaction further downstream. Air and fuel velocities and fuel drop size results are also reported.

  17. The use of experimental design to find the operating maximum power point of PEM fuel cells

    SciTech Connect

    Crăciunescu, Aurelian; Pătularu, Laurenţiu; Ciumbulea, Gloria; Olteanu, Valentin; Pitorac, Cristina; Drugan, Elena

    2015-03-10

    Proton Exchange Membrane (PEM) Fuel Cells are difficult to model due to their complex nonlinear nature. In this paper, the development of a PEM Fuel Cells mathematical model based on the Design of Experiment methodology is described. The Design of Experiment provides a very efficient methodology to obtain a mathematical model for the studied multivariable system with only a few experiments. The obtained results can be used for optimization and control of the PEM Fuel Cells systems.

  18. A systematic experimental and computational investigation of a class of contoured wall fuel injectors

    NASA Technical Reports Server (NTRS)

    Waitz, Ian A.; Marble, Frank E.; Zukoski, Edward E.

    1992-01-01

    The performance of contoured wall fuel injectors for scramjet engine applications is considered. These fuel injectors were aimed at augmenting mixing through axial vorticity production arising from interaction of the fuel/air interface with an oblique shock. The effects of incoming boundary layer height, injector spacing, and injectant to freestream pressure and velocity ratios are examined. Results from 3D flow field surveys and Navier-Stokes simulations are presented.

  19. High energy X-ray diffraction measurement of residual stresses in a monolithic aluminum clad uranium–10 wt% molybdenum fuel plate assembly

    SciTech Connect

    D. W. Brown; M. A. Okuniewski; J. D. Almer; L. Balogh; B. Clausen; J. S. Okasinski; B. H. Rabin

    2013-10-01

    Residual stresses are