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Sample records for actual fuel rod

  1. 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.

  2. FUEL ROD CLUSTERS

    DOEpatents

    Schultz, A.B.

    1959-08-01

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

  3. Locked-wrap fuel rod

    DOEpatents

    Kaplan, Samuel; Chertock, Alan J.; Punches, James R.

    1977-01-01

    A method for spacing fast reactor fuel rods using a wire wrapper improved by orienting the wire-wrapped fuel rods in a unique manner which introduces desirable performance characteristics not attainable by previous wire-wrapped designs. Use of this method in a liquid metal fast breeder reactor results in: (a) improved mechanical performance, (b) improved rod-to-rod contact, (c) reduced steel volume, and (d) improved thermal-hydraulic performance. The method produces a "locked wrap" design which tends to lock the rods together at each of the wire cluster locations.

  4. Stuck fuel rod capping sleeve

    DOEpatents

    Gorscak, Donald A.; Maringo, John J.; Nilsen, Roy J.

    1988-01-01

    A stuck fuel rod capping sleeve to be used during derodding of spent fuel assemblies if a fuel rod becomes stuck in a partially withdrawn position and, thus, has to be severed. The capping sleeve has an inner sleeve made of a lower work hardening highly ductile material (e.g., Inconel 600) and an outer sleeve made of a moderately ductile material (e.g., 304 stainless steel). The inner sleeve may be made of an epoxy filler. The capping sleeve is placed on a fuel rod which is then severed by using a bolt cutter device. Upon cutting, the capping sleeve deforms in such a manner as to prevent the gross release of radioactive fuel material

  5. 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.

  6. Tribe kills fuel rod proposal

    SciTech Connect

    1995-02-13

    This article is a review of nuclear utilities` efforts to find a repository of spent fuel rods. The rejection by the Mescalero Apaches of plans to build a waste repository on tribal lands has left a number of utilities scrambling to find interim solutions. Prairie Island will have to close before the end of the year unless a solution is found, and the Hope Creek/Salem units, exhausting there storage capacity within ten years, are considering dry-cask storage.

  7. Fuel followed control rod installation at AFRRI

    SciTech Connect

    Moore, Mark; Owens, Chris; Forsbacka, Matt

    1992-07-01

    Fuel Followed Control Rods (FFCRs) were installed at the Armed Forces Radiobiology Research Institute's 1 MW TRIGA Reactor. The procedures for obtaining, shipping, and installing the FFCRs is described. As part of the FFCR installation, the transient rod drive was relocated. Core performance due to the addition of the fuel followed control rods is discussed. (author)

  8. Nuclear design of Helical Cruciform Fuel rods

    SciTech Connect

    Shirvan, K.; Kazimi, M. S.

    2012-07-01

    In order to increase the power density of current and new light water reactor designs, the Helical Cruciform Fuel (HCF) rods are proposed. The HCF rods are equivalent to a cylindrical rod, with the fuel in a cruciform shaped, twisted axially. The HCF rods increase the surface area to volume ratio and inter-subchannel mixing behavior due to their cruciform and helical shapes, respectively. In a previous study, the HCF rods have shown the potential to up-rate existing PWRs by 50% and BWRs by 25%. However, HCF rods do display different neutronics modeling and performance. The cruciform cross section of HCF rods creates radially asymmetric heat generation and temperature distribution. The nominal HCF rod's beginning of life reactivity is reduced, compared to a cylindrical rod with the same fuel volume, by 500 pcm, due to increase in absorption in cladding. The rotation of these rods accounts for reactivity changes, which depends on the H/HM ratio of the pin cell. The HCF geometry shows large sensitivities to U{sup 235} or gadolinium enrichments compared to a cylindrical geometry. In addition, the gadolinium-containing HCF rods show a stronger effect on neighboring HCF rods than in case of cylindrical rods, depending on the orientation of the HCF rods. The helical geometry of the rods introduces axial shadowing of about 600 pcm, not seen in typical cylindrical rods. (authors)

  9. Double-clad nuclear fuel safety rod

    DOEpatents

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  10. Fuel axial relocation in ballooning fuel rods. [PWR; BWR

    SciTech Connect

    Siefken, L.J.

    1983-01-01

    Fuel movement, in the longitudinal direction in ballooning fuel rods, shifts the position of heat generation and may cause an increase in cladding temperature in the ballooning region. This paper summarizes the axial fuel relocation data obtained in fuel rod tests conducted in the United States and the Federal Republic of Germany, describes a model for calculating fuel axial relocation, and gives a quantitative analysis of the impact of fuel relocation on cladding temperature. The amount of fuel relocation in 18 ballooned fuel rods was determined from neutron radiographs, niobium gamma decay counts, and photomicrographs. The fuel rods had burnups in the range of 0 to 35,000 MWd/t and cladding hoop strains varying from 0 to 72%.

  11. International symposium on fuel rod simulators: development and application

    SciTech Connect

    McCulloch, R.W.

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  12. Analysis of Double-encapsulated Fuel Rods

    SciTech Connect

    Hales, Jason Dean; Medvedev, Pavel G; Novascone, Stephen Rhead; Perez, Danielle Marie; Williamson, Richard L

    2014-09-01

    In an LWR fuel rod, the cladding encapsulates the fuel, contains fission products, and transfers heat directly to the water coolant. In some situations, it may be advantageous to separate the cladding from the coolant through use of a secondary cladding or capsule. This may be done to increase confidence that the fuel or fission products will not mix with the coolant, to provide a mechanism for controlling the rod temperature, or to place multiple experimental rodlets within a single housing. With an axisymmetric assumption, it is possible to derive closed-form expressions for the temperature profile in a fuel rod using radially-constant thermal conductivity in the fuel. This is true for both a traditional fuel-cladding rod and a double-encapsulated fuel (fuel, cladding, capsule) configuration. Likewise, it is possible to employ a fuel performance code to analyse both a traditional and a double-encapsulated fuel. In the case of the latter, two sets of gap heat transfer conditions must be imposed. In this work, we review the equations associated with radial heat transfer in a cylindrical system, present analytic and computational results for a postulated power and gas mixture history for IFA-744, and describe the analysis of the AFC-2A, 2B metallic fuel alloy experiments at the Advanced Test Reactor, including the effect of a release of fission products into the cladding-capsule gap. The computational results for these two cases were obtained using BISON, a fuel performance code under development at Idaho National Laboratory.

  13. Nuclear reactor fuel rod attachment system

    DOEpatents

    Not Available

    1980-09-17

    A reusable system is described for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member. The locking cap has two opposing fingers shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed.

  14. Method and means of packaging nuclear fuel rods for handling

    DOEpatents

    Adam, Milton F.

    1979-01-01

    Nuclear fuel rods, especially spent nuclear fuel rods that may show physical distortion, are encased within a metallic enclosing structure by forming a tube about the fuel rod. The tube has previously been rolled to form an overlapping tubular structure and then unrolled and coiled about an axis perpendicular to the tube. The fuel rod is inserted into the tube as the rolled tube is removed from a coiled strip and allowed to reassume its tubular shape about the fuel rod. Rollers support the coiled strip in an open position as the coiled strip is uncoiled and allowed to roll about the fuel rod.

  15. Grid-to-rod flow-induced impact study for PWR fuel in reactor

    DOE PAGESBeta

    Jiang, Hao; Qu, Jun; Lu, Roger Y.; Wang, Jy-An John

    2016-06-10

    The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effectivemore » way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.« less

  16. Nuclear reactor fuel rod attachment system

    DOEpatents

    Christiansen, David W.

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

  17. LWR fuel rod bundle behavior under severe fuel damage conditions

    SciTech Connect

    Kuczera, B. Hagen, S.; Hofmann, P.

    1988-01-01

    Light water reactor (LWR) safety research and development activities conducted at Kernforschungszentrum Karlsruhe have recently been reorganized with a concentrated mission under the LWR safety project group. The topics treated relate mainly to severe-accident analysis research and source term assessment as well as to source term mitigation measures. A major part of the investigations concerns the early phase of a severe core meltdown accident, specifically LWR rod assembly behavior under sever fuel damage (SFD) conditions. To determine the extent of fuel rod damage, including the relocation behavior of molten reaction products, damage propagation, time-dependent H{sub 2} generation from clad oxidation, and fragmentation of oxygen-embrittled materials during cooldown and quenching, extensive out-of-pile rod bundle experiments have been initiated in the new CORA test facility. The bundle parameters, such as rod dimensions, rod pitch, and grid spacer, can be adjusted to both pressurized water reactor (PWR) and boiling water reactor (BWR) conditions. Currently, the test program consists of 15 experiments in which the influence of Inconel grid spacer, (Ag,In,Cd)-absorber rods (PWR) and of B{sub 4}C control blades (BWR) on fuel damage initiation and damage propagation are being investigated for different boundary conditions. As of June 1988, four bundle tests had been successfully carried out for PWR accident conditions.

  18. Electric Fuel Rod Simulator Fabrication at ORNL

    SciTech Connect

    Ott, Larry J.; McCulloch, Reg

    2004-02-04

    Commercial vendors could not supply the high-quality, highly instrumented electric fuel rod simulators (FRS) required for large thermal-hydraulic safety-oriented experiments at the Oak Ridge National Laboratory (ORNL) in the 1970s and early 1980s. Staff at ORNL designed, developed, and manufactured the simulators utilized in these safety experiments. Important FRS design requirements include (1) materials of construction, (2) test power requirements and availability, (3) experimental test objectives, (4) supporting thermal analyses, and (5) extensive quality control throughout all phases of FRS fabrication. This paper will present an overview of these requirements (design, analytics, and quality control) as practiced at ORNL to produce a durable high-quality FRS.

  19. Electric Fuel Rod Simulator Fabrication at ORNL

    NASA Astrophysics Data System (ADS)

    Ott, Larry J.; McCulloch, Reg

    2004-02-01

    Commercial vendors could not supply the high-quality, highly instrumented electric fuel rod simulators (FRS) required for large thermal-hydraulic safety-oriented experiments at the Oak Ridge National Laboratory (ORNL) in the 1970s and early 1980s. Staff at ORNL designed, developed, and manufactured the simulators utilized in these safety experiments. Important FRS design requirements include (1) materials of construction, (2) test power requirements and availability, (3) experimental test objectives, (4) supporting thermal analyses, and (5) extensive quality control throughout all phases of FRS fabrication. This paper will present an overview of these requirements (design, analytics, and quality control) as practiced at ORNL to produce a durable high-quality FRS.

  20. System for fuel rod removal from a reactor module

    DOEpatents

    Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

    1988-07-28

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

  1. System for fuel rod removal from a reactor module

    DOEpatents

    Matchett, Richard L.; Roof, David R.; Kikta, Thomas J.; Wilczynski, Rosemarie; Nilsen, Roy J.; Bacvinskas, William S.; Fodor, George

    1990-01-01

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system.

  2. Fuel rod mechanical deformation during the PBF/LOFT lead rod loss-of-coolant experiments

    SciTech Connect

    Varacalle, Jr., D. J.; MacDonald, P. E.; Shiozawa, S.; Driskell, W. E.

    1980-01-01

    Results of four PBF/LOFT Lead Rod (LLR) sequential blowdown tests conducted in the Power Burst Facility (PBF) are presented. Each test employed four separately shrouded fuel rods. The primary objective of the test series was to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa), light water reactor design fuel rods when subjected to a series of double ended cold leg break loss-of-coolant accident (LOCA) tests, and to determine whether subjecting these deformed fuel rods to subsequent testing would result in rod failure. The extent of mechanical deformation (buckling, collapse, or waisting of the cladding) was evaluated by comparison of cladding temperature and system pressure measurements with out-of-pile experimental data, and by posttest visual examinations and cladding diametral measurements.

  3. Experimental study of burnout in channels with twisted fuel rods

    NASA Astrophysics Data System (ADS)

    Bol'Shakov, V. V.; Bashkirtsev, S. M.; Kobzar', L. L.; Morozov, A. G.

    2007-05-01

    The results of experimental studies of pressure drop and critical heat flux in the models of fuel assemblies (FAs) with fuel rod simulators twisted relative to the longitudinal axis and a three-ray cross section are considered. The experimental data are compared to the results obtained with the use of techniques adopted for design calculations with fuel rod bundles of type-VVER reactors.

  4. BWR fuel rod performance evaluation program. Final report

    SciTech Connect

    Rowland, T.C.

    1986-05-01

    The joint EPRI/GE fuel performance program, RP510-1, involved thorough preirradiation characterization of fuel used in lead test assemblies, detailed surveillance of their operation, and interim site examinations of the assemblies during reactor outages. The program originally included four GE-5 lead test assemblies operating in the Peach Bottom-2 reactor. The program was later modified to include the pressurized fuel rod test assembly in the Peach Bottom-3 reactor (RP510-2). The program modification also included extending the operation of the Peach Bottom-2 and Peach Bottom-3 lead test assembly fuel beyond normal discharge exposures. Interim site examination results following the first four cycles of operation of the Peach Bottom-2 lead test assemblies up to 35 GWd/MT and the examination of the Peach Bottom-3 pressurized test assembly at 32 GWd/MT are presented in this report. Elements of the examinations included visual examination of the fuel bundles; individual fuel rod visual examinations, rod length measurements, ultrasonic and eddy current nondestructive testing, Zircaloy cladding oxide thickness measurements and fission gas measurements. Channel measurements were made on the PB-2 Lead Test Assemblies after each of the first three operating cycles. All of the bundles were found to be in good condition. Since the pressurized test assembly contained pressurized and nonpressurized fuel rods in symmetric positions, it was possible to make direct comparisons of the fission gas release from pairs of pressurized and nonpressurized fuel rods with identical power histories. With one exception, the release was less from the pressurized fuel rod of each pair. Fuel rod power histories were calculated using new physics methods for all of the fuel rods that were punctured for fission gas release measurements. 28 refs., 41 figs., 16 tabs.

  5. Apparatus for injection casting metallic nuclear energy fuel rods

    DOEpatents

    Seidel, Bobby R.; Tracy, Donald B.; Griffiths, Vernon

    1991-01-01

    Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

  6. Double-clad nuclear-fuel safety rod

    DOEpatents

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  7. End-of-life nondestructive examination of Light Water Breeder Reactor fuel rods (LWBR Development Program)

    SciTech Connect

    Gorscak, D.A.; Campbell, W.R.; Clayton, J.C.

    1987-10-01

    In-bundle and out-of-bundle (single rod) nondestructive examinations of Light Water Breeder Reactor fuel rods were performed. In-bundle examinations included visual examination and measurement of rod bow, rod-to-rod gaps, and rod removal forces. Out-of-bundle examinations included rod visuals and measurement of fuel rod length, diameter and ovality, cladding oxide and crud thickness, support grid induced cladding wear mark depth and volume, and fuel rod free hanging bow. The out-of-bundle examination also included ultrasonic inspection for cladding defects, neutron radiography for pellet integrity and plenum gap measurements, and gamma scans for instack axial gap screening and binary fuel stack length measurements. The measurements confirmed design predictions of fuel rod performance and provided evidence of excellent fuel rod performance for operation of Light Water Breeder Reactor to 29,047 effective full power hours (EFPH).

  8. High burnup effects in WWER fuel rods

    SciTech Connect

    Smirnov, V.; Smirnov, A.

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

  9. Fretting wear behaviors of a dual-cooled nuclear fuel rod under a simulated rod vibration

    SciTech Connect

    Lee, Young-Ho; Kim, Hyung-Kyu; Kang, Heung-Seok; Yoon, Kyung-Ho; Kim, Jae-Yong; Lee, Kang-Hee

    2012-06-06

    Recently, a dual-cooled fuel (i.e., annular fuel) that is compatible with current operating PWR plants has been proposed in order to realize both a considerable amount of power uprating and an increase of safety margins. As the design concept should be compatible with current operating PWR plants, however, it shows a narrow gap between the fuel rods when compared with current solid nuclear fuel arrays and needs to modify the spacer grid shapes and their positions. In this study, fretting wear tests have been performed to evaluate the wear resistance of a dual-cooled fuel by using a proposed spring and dimple of spacer grids that have a cantilever type and hemispherical shape, respectively. As a result, the wear volume of the spring specimen gradually increases as the contact condition is changed from a certain gap, just contact to positive force. However, in the dimple specimen, just contact condition shows a large wear volume. In addition, a circular rod motion at upper region of contact surface is gradually increased and its diametric size depends on the wear depth increase. Based on the test results, the fretting wear resistance of the proposed spring and dimple is analyzed by comparing the wear measurement results and rod motion in detail.

  10. Fretting wear behaviors of a dual-cooled nuclear fuel rod under a simulated rod vibration

    NASA Astrophysics Data System (ADS)

    Lee, Young-Ho; Kim, Hyung-Kyu; Kang, Heung-Seok; Yoon, Kyung-Ho; Kim, Jae-Yong; Lee, Kang-Hee

    2012-06-01

    Recently, a dual-cooled fuel (i.e., annular fuel) that is compatible with current operating PWR plants has been proposed in order to realize both a considerable amount of power uprating and an increase of safety margins. As the design concept should be compatible with current operating PWR plants, however, it shows a narrow gap between the fuel rods when compared with current solid nuclear fuel arrays and needs to modify the spacer grid shapes and their positions. In this study, fretting wear tests have been performed to evaluate the wear resistance of a dual-cooled fuel by using a proposed spring and dimple of spacer grids that have a cantilever type and hemispherical shape, respectively. As a result, the wear volume of the spring specimen gradually increases as the contact condition is changed from a certain gap, just contact to positive force. However, in the dimple specimen, just contact condition shows a large wear volume. In addition, a circular rod motion at upper region of contact surface is gradually increased and its diametric size depends on the wear depth increase. Based on the test results, the fretting wear resistance of the proposed spring and dimple is analyzed by comparing the wear measurement results and rod motion in detail.

  11. Investigation of Backscatter X-ray imaging techniques for Uranium Dioxide Fuel Rods

    SciTech Connect

    Jackson, Timothy D; Hollenbach, Daniel F; Shedlock, Daniel

    2011-01-01

    Radiography by Selective Detection (RSD), was investigated for its ability to determine the presence and types of defects in a UO{sub 2} fuel rod surrounded by zirconium cladding. Images created using a Monte Carlo model compared favorably with actual X-ray backscatter images from mock fuel rods. A fuel rod was modeled as a rectangular parallelepiped with zirconium cladding, and pencil beam X-ray sources of 160 kVp (79 keV avg) and 480 kVp (218 keV avg) were generated using the Monte Carlo N-Particle Transport Code to attempt to image void and palladium (Pd) defects in the interior and on the surface of the fuel pellet. It was found that the 160 kVp spectrum was unable to detect the presence of interior defects, whereas the 480 kVp spectrum detected them with both the standard and the RSD backscatter methods, though the RSD method was very inefficient. It was also found that both energy spectra were able to detect void and Pd defects on the surface using both imaging methods. Additionally, two mock fuel rods were imaged using a backscatter X-ray imaging system, one consisting of hafnium pellets in a Zircaloy-4 cladding and the other consisting of steel pellets in a Zircalloy-4 cladding which was then encased in a steel cladding (a double encapsulation configuration employed in irradiation and experiments). It was found that the system was capable of detecting individual HfO{sub 2} pellets in a Zircaloy-4 cladding and may be capable of detecting individual steel pellets in the double-encapsulated sample. It is expected that the system would also be capable of detecting individual UO{sub 2} pellets in a Zircaloy-4 cladding, though no UO{sub 2} fuel rod was available for imaging.

  12. Process development and fabrication for sphere-pac fuel rods. [PWR; BWR

    SciTech Connect

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted.

  13. Overview of Fuel Rod Simulator Usage at ORNL

    SciTech Connect

    Ott, Larry J.; McCulloch, Reg

    2004-02-04

    During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

  14. Overview of Fuel Rod Simulator Usage at ORNL

    NASA Astrophysics Data System (ADS)

    Ott, Larry J.; McCulloch, Reg

    2004-02-01

    During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

  15. Code System to Calculate Fuel Rod Thermal Performance.

    2000-11-27

    Version: 00 GT2R2 is Revision 2 of GAPCON-THERMAL-2 and is used to calculate the thermal behavior of a nuclear fuel rod during normal steady-state operation. The program was developed as a tool for estimating fuel-cladding gap conductances and fuel-stored energy. Models used include power history, fission gas generation and release, fuel relocation and densification, and fuel-cladding gap conductance. The gas release and relocation models can be used to make either best-estimate or conservative predictions. Themore » code is used by the United States Nuclear Regulatory Commission for audit calculations of nuclear fuel thermal performance computer codes.« less

  16. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    SciTech Connect

    Marschman, Steven C.; Warmann, Stephan A.; Rusch, Chris

    2014-03-01

    , inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage. To document the initial condition of the used fuel prior to emplacement in a storage system, “sister ” fuel rods will be harvested and sent to a national laboratory for characterization and archival purposes. This report supports the demonstration by describing how sister rods will be shipped and received at a national laboratory, and recommending basic nondestructive and destructive analyses to assure the fuel rods are adequately characterized for UFDC work. For this report, a hub-and-spoke model is proposed, with one location serving as the hub for fuel rod receipt and characterization. In this model, fuel and/or clad would be sent to other locations when capabilities at the hub were inadequate or nonexistent. This model has been proposed to reduce DOE-NE’s obligation for waste cleanup and decontamination of equipment.

  17. Optical coherence tomography for nondestructive evaluation of fuel rod degradation

    SciTech Connect

    Renshaw, Jeremy B.; Jenkins, Thomas P. Buckner, Benjamin D.; Friend, Brian

    2015-03-31

    Nuclear power plants regularly inspect fuel rods to ensure safe and reliable operation. Excessive corrosion can cause fuel failures which can have significant repercussions for the plant, including impacts on plant operation, worker exposure to radiation, and the plant's INPO rating. While plants typically inspect for fuel rod corrosion using eddy current techniques, these techniques have known issues with reliability in the presence of tenacious, ferromagnetic crud layers that can deposit during operation, and the nondestructive evaluation (NDE) inspection results can often be in error by a factor of 2 or 3. For this reason, alternative measurement techniques, such as Optical Coherence Tomography (OCT), have been evaluated that are not sensitive to the ferromagnetic nature of the crud. This paper demonstrates that OCT has significant potential to characterize the thickness of crud layers that can deposit on the surfaces of fuel rods during operation. Physical trials have been performed on simulated crud samples, and the resulting data show an apparent correlation between the crud layer thickness and the OCT signal.

  18. Optical coherence tomography for nondestructive evaluation of fuel rod degradation

    NASA Astrophysics Data System (ADS)

    Renshaw, Jeremy B.; Jenkins, Thomas P.; Buckner, Benjamin D.; Friend, Brian

    2015-03-01

    Nuclear power plants regularly inspect fuel rods to ensure safe and reliable operation. Excessive corrosion can cause fuel failures which can have significant repercussions for the plant, including impacts on plant operation, worker exposure to radiation, and the plant's INPO rating. While plants typically inspect for fuel rod corrosion using eddy current techniques, these techniques have known issues with reliability in the presence of tenacious, ferromagnetic crud layers that can deposit during operation, and the nondestructive evaluation (NDE) inspection results can often be in error by a factor of 2 or 3. For this reason, alternative measurement techniques, such as Optical Coherence Tomography (OCT), have been evaluated that are not sensitive to the ferromagnetic nature of the crud. This paper demonstrates that OCT has significant potential to characterize the thickness of crud layers that can deposit on the surfaces of fuel rods during operation. Physical trials have been performed on simulated crud samples, and the resulting data show an apparent correlation between the crud layer thickness and the OCT signal.

  19. Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

    DOE PAGESBeta

    Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; Nadiga, Balasubramanya T.; Karoutas, Zeses; Berndt, Markus

    2016-06-29

    Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

  20. Control rod system useable for fuel handling in a gas-cooled nuclear reactor

    DOEpatents

    Spurrier, Francis R.

    1976-11-30

    A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

  1. Electric heater for nuclear fuel rod simulators

    DOEpatents

    McCulloch, Reginald W.; Morgan, Jr., Chester S.; Dial, Ralph E.

    1982-01-01

    The present invention is directed to an electric cartridge-type heater for use as a simulator for a nuclear fuel pin in reactor studies. The heater comprises an elongated cylindrical housing containing a longitudinally extending helically wound heating element with the heating element radially inwardly separated from the housing. Crushed cold-pressed preforms of boron nitride electrically insulate the heating element from the housing while providing good thermal conductivity. Crushed cold-pressed preforms of magnesia or a magnesia-15 percent boron nitride mixture are disposed in the cavity of the helical heating element. The coefficient of thermal expansion of the magnesia or the magnesia-boron nitride mixture is higher than that of the boron nitride disposed about the heating element for urging the boron nitride radially outwardly against the housing during elevated temperatures to assure adequate thermal contact between the housing and the boron nitride.

  2. Preliminary design report for the prototypical fuel rod consolidation system

    SciTech Connect

    Rosa, J.M.

    1986-12-03

    This report documents NUTECH's preliminary design of a dry, spent fuel rod consolidation system. This preliminary design is the result of Phase I of a planned four phase project. The present report on this project provides a considerable amount of detail for a preliminary design effort. The design and all of its details are described in this Preliminary Design Report (PDR). The NUTECH dry rod consolidation system described herein is remotely operated. It provides for automatic operation, but with operator hold points between key steps in the process. The operator has the ability to switch to a manual operation mode at any point in the process. The system is directed by the operator using an executive computer which controls and coordinates the operation of the in-cell equipment. The operator monitors the process using an in-cell closed circuit television (CCTV) system with audio output and equipment status displays on the computer monitor. The in-cell mechanical equipment consists of the following: (1) two overhead cranes with manipulators; (2) a multi-degree of freedom fuel handling table and its clamping equipment; (3) a fuel assembly end fitting removal station and its tools; (4) a consolidator (which pulls rods, assembles the consolidated bundle and loads the canister); (5) a canister end cap welder and weld inspection system; (6) decontamination systems; and (7) the CCTV and microphone systems.

  3. The analysis of failed nuclear fuel rods by gamma computed tomography

    NASA Astrophysics Data System (ADS)

    Dobrin, Relu; Craciunescu, Teddy; Tuturici, Ioan Liviu

    1997-07-01

    The failure of the cladding of an irradiated nuclear fuel rod can lead to the loss of some γ-radioactive fission products. Consequently the distribution of these fission products is altered in the cross-section of the fuel rod. The modification of the distribution, obtained by gamma computed tomography, is used to determine the integrity of the fuel cladding. The paper reports an experimental result, obtained for a CANDU-type fuel rod, irradiated in a TRIGA 14 MWth reactor.

  4. Multidimensional simulations of hydrides during fuel rod lifecycle

    NASA Astrophysics Data System (ADS)

    Stafford, D. S.

    2015-11-01

    In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim.

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

    SciTech Connect

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

    1993-03-01

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

  6. Full-length fuel rod behavior under severe accident conditions

    SciTech Connect

    Lombardo, N J; Lanning, D D; Panisko, F E

    1992-12-01

    This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors.

  7. 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.

  8. 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.

  9. 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.

  10. Automatic inspection system for nuclear fuel pellets or rods

    DOEpatents

    Miller, Jr., William H.; Sease, John D.; Hamel, William R.; Bradley, Ronnie A.

    1978-01-01

    An automatic inspection system is provided for determining surface defects on cylindrical objects such as nuclear fuel pellets or rods. The active element of the system is a compound ring having a plurality of pneumatic jet units directed into a central bore. These jet units are connected to provide multiple circuits, each circuit being provided with a pressure sensor. The outputs of the sensors are fed to a comparator circuit whereby a signal is generated when the difference of pressure between pneumatic circuits, caused by a defect, exceeds a pre-set amount. This signal may be used to divert the piece being inspected into a "reject" storage bin or the like.

  11. ADS-Demo Fuel Rod Performance: Multivariate Statistical Analysis

    SciTech Connect

    Calabrese, R.; Vettraino, F.; Luzzi, L.

    2004-07-01

    A forward step in the development of Accelerator Driven System (ADS) for the Pu, MA and LLFP transmutation, is the realisation of a 80 MWt ADS-demo (XADS) whose basic objective is the system feasibility demonstration. The XADS is forecasted to adopt the UO{sub 2}-PuO{sub 2} mixed-oxides fuel already experimented in the sodium cooled fast reactors such as the french SPX-1. The present multivariate statistical analysis performed by using the Transuranus Code, was carried out for the Normal Operation at the so-called Enhanced Nominal Conditions (120% nominal reactor power), aimed at verifying that the fuel system complies with the stated design limits, i.e. centerline fuel temperature, cladding temperature and damage, during all the in-reactor lifetime. A statistical input set similar to SPX and PEC fuel case, was adopted. One most relevant assumption in the present calculations was a 30% AISI-316 cladding thickness corrosion at EOL. Relative influence of main fuel rod parameters on fuel centerline temperature was also evaluated. (authors)

  12. Final Report: Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation

    SciTech Connect

    Rowsell, David Leon

    2015-06-01

    This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

  13. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    NASA Astrophysics Data System (ADS)

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

    1991-02-01

    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  14. Experimental evaluation of cobalt behavior on BWR fuel rod surface

    SciTech Connect

    Karasawa, H.; Asakura, Y.; Sakagami, M.; Uchida, S. )

    1988-06-01

    Cobalt behavior on the boiling water reactor (BWR) fuel rod surface was experimentally evaluated at 285 C and with various pH values. Adsorption of cobalt ions on hematite particles proceeded via the exchange reaction of cobalt ion with the surface hydroxyl of the hematite. The equilibrium constant for the adsorption at 285 C was found to be -- 570 times as large as that at 20 C. The adsorbate formed cobalt ferrite at the rate of 3.4 x 10/sup -2/ g-Co/g-Co adsorbed/h. The dissolution rates of cobalt ferrite and cobalt oxide particles were found to depend on (H/sup -/)/sup 1.1/ and (H/sup -/)/sup 1.2/, respectively, where (H/sup -/) means the H/sup -/ concentration. Cobalt ions were released from these oxides when O/sup 2-/ ions in them combined with two aqueous protons to form water at the oxide-water interface. Cobalt behavior on the fuel rod surface under BWR conditions was discussed using the experimental results.

  15. Fission gas release from UO{sub 2+x} in defective light water reactor fuel rods

    SciTech Connect

    Skim, Y. S.

    1999-11-12

    A simplified semi-empirical model predicting fission gas release form UO{sub 2+x} fuel to the fuel rod plenum as a function of stoichiometry excess (x) is developed to apply to the fuel of a defective LWR fuel rod in operation. The effect of fuel oxidation in enhancing gas diffusion is included as a parabolic dependence of the stoichiometry excess. The increase of fission gas release in a defective BWR fuel rod is at the most 3 times higher than in an intact fuel rod because of small extent of UO{sub 2} oxidation. The major enhancement contributor in fission gas release of UO{sub 2+x} fuel is the increased diffusivity due to stoichiometry excess rather than the higher temperature caused by degraded fuel thermal conductivity.

  16. A Neural Network Model for the Tomographic Analysis of Irradiated Nuclear Fuel Rods

    SciTech Connect

    Craciunescu, Teddy

    2004-04-15

    A tomographic method based on a multilayer feed-forward artificial neural network is proposed for the reconstruction of gamma-radioactive fission product distribution in irradiated nuclear fuel rods. The quality of the method is investigated as compared to a conventional technique on experimental results concerning a Canada deuterium uranium reactor (CANDU)-type fuel rod irradiated in a TRIGA reactor.

  17. Destructive examination of 3-cycle LWR fuel rods from Turkey Point Unit 3 for the Climax-Spent Fuel Test

    SciTech Connect

    Atkin, S.D.

    1981-06-01

    The destructive examination results of five light water reactor rods from the Turkey Point Unit 3 reactor are presented. The examinations included fission gas collection and analyses, burnup and hydrogen analyses, and a metallographic evaluation of the fuel, cladding, oxide, and hydrides. The rods exhibited a low fission gas release with all other results appearing representative for pressurized water reactor fuel rods with similar burnups (28 GWd/MTU) and operating histories.

  18. Development of burnup dependent fuel rod model in COBRA-TF

    NASA Astrophysics Data System (ADS)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  19. A Multi-Stage Wear Model for Grid-to-Rod Fretting of Nuclear Fuel Rods

    SciTech Connect

    Blau, Peter Julian

    2014-01-01

    The wear of fuel rod cladding against the supporting structures in the cores of pressurized water nuclear reactors (PWRs) is an important and potentially costly tribological issue. Grid-to-rod fretting (GTRF), as it is known, involves not only time-varying contact conditions, but also elevated temperatures, flowing hot water, aqueous tribo-corrosion, and the embrittling effects of neutron fluences. The multi-stage, closed-form analytical model described in this paper relies on published out-of-reactor wear and corrosion data and a set of simplifying assumptions to portray the conversion of frictional work into wear depth. The cladding material of interest is a zirconium-based alloy called Zircaloy-4, and the grid support is made of a harder and more wear-resistant material. Focus is on the wear of the cladding. The model involves an incubation stage, a surface oxide wear stage, and a base alloy wear stage. The wear coefficient, which is a measure of the efficiency of conversion of frictional work into wear damage, can change to reflect the evolving metallurgical condition of the alloy. Wear coefficients for Zircaloy-4 and for a polyphase zirconia layer were back-calculated for a range of times required to wear to a critical depth. Inputs for the model, like the friction coefficient, are taken from the tribology literature in lieu of in-reactor tribological data. Concepts of classical fretting were used as a basis, but are modified to enable the model to accommodate the complexities of the PWR environment. Factors like grid spring relaxation, pre-oxidation of the cladding, multiple oxide phases, gap formation, impact, and hydrogen embrittlement are part of the problem definition but uncertainties in their relative roles limits the ability to validate the model. Sample calculations of wear depth versus time in the cladding illustrate how GTRF wear might occur in a discontinuous fashion during months-long reactor operating cycles. A means to account for grid/rod gaps

  20. Interaction of Failed Fuel Rods Under Air Ingress Conditions

    SciTech Connect

    Hozer, Zoltan; Windberg, Peter; Nagy, Imre; Maroti, Laszlo; Matus, Lajos; Horvath, Marta; Csordas, Anna Pinter; Balasko, Marton; Czitrovszky, Aladar; Jani, Peter

    2003-03-15

    In the late phase of a severe reactor accident, the molten corium interacts with the vessel wall, and it can lead to the failure of the lower head. Through the failed bottom wall, part of the corium can flow into the cavity, and air can enter the primary circuit. The residual fuel in the core periphery will be further oxidized in air atmosphere. The degradation process will accelerate, and new chemical species will be formed, which can have an impact on the release of radioactive materials.Two experiments were carried out with electrically heated nine-rod pressurized water reactor-type bundles in the CODEX (COre Degradation EXperiment) facility to provide experimental data on the behavior of real fuel bundles under air oxidation conditions. The main objective of the tests was the investigation of oxidation phenomena, and some other important aspects (e.g., enhanced fission product release) were not addressed.The CODEX air ingress tests indicated the acceleration of oxidation phenomena and core degradation processes during the late phase of the vessel melt through accident, when air can have access to the residual fuel bundles in the reactor core. The degradation process was accompanied with zirconium-nitride formation and release of uranium-rich aerosols.

  1. Uncertainty analysis of spent nuclear fuel isotopics and rod internal pressure

    NASA Astrophysics Data System (ADS)

    Bratton, Ryan N.

    The bias and uncertainty in fuel isotopic calculations for a well-defined radio- chemical assay benchmark are investigated with Sampler, the new sampling-based uncertainty quantification tool in the SCALE code system. Isotopic predictions are compared to measurements of fuel rod MKP109 of assembly D047 from the Calvert Cliffs Unit 1 core at three axial locations, representing a range of discharged fuel burnups. A methodology is developed which quantifies the significance of input parameter uncertainties and modeling decisions on isotopic prediction by compar- ing to isotopic measurement uncertainties. The SCALE Sampler model of the D047 assembly incorporates input parameter uncertainties for key input data such as multigroup cross sections, decay constants, fission product yields, the cladding thickness, and the power history for fuel rod MKP109. The effects of each set of input parameter uncertainty on the uncertainty of isotopic predictions have been quantified. In this work, isotopic prediction biases are identified and an investiga- tion into their sources is proposed; namely, biases have been identified for certain plutonium, europium, and gadolinium isotopes for all three axial locations. More- over, isotopic prediction uncertainty resulting from only nuclear data is found to be greatest for Eu-154, Gd-154, and Gd-160. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle as- sembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each considered WBN1 fuel rod. An alternate model for the amount of helium released from zirconium diboride (ZrB2) integral fuel burn- able absorber (IFBA) layers is

  2. Core design study of a supercritical light water reactor with double row fuel rods

    SciTech Connect

    Zhao, C.; Wu, H.; Cao, L.; Zheng, Y.; Yang, J.; Zhang, Y.

    2012-07-01

    An equilibrium core for supercritical light water reactor has been designed. A novel type of fuel assembly with dual rows of fuel rods between water rods is chosen and optimized to get more uniform assembly power distributions. Stainless steel is used for fuel rod cladding and structural material. Honeycomb structure filled with thermal isolation is introduced to reduce the usage of stainless steel and to keep moderator temperature below the pseudo critical temperature. Water flow scheme with ascending coolant flow in inner regions is carried out to achieve high outlet temperature. In order to enhance coolant outlet temperature, the radial power distributions needs to be as flat as possible through operation cycle. Fuel loading pattern and control rod pattern are optimized to flatten power distribution at inner regions. Axial fuel enrichment is divided into three parts to control axial power peak, which affects maximum cladding surface temperature. (authors)

  3. Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods (LWBR Development Program)

    SciTech Connect

    Clayton, J.C.

    1985-08-01

    Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxide thickness of non-defected rods, gave results which were in reasonable agreement with the outer surface oxide thicknesses of defected rods. When the analysis procedure was modified to account for additional corrosion proportional to fission rate and to time, the calculated values agreed well with measured inner oxide corrosion film values. Hydrogen pickup in the defected rods was not directly proportional to local corrosion oxide weight gain as was the case for non-defected rods. 16 refs., 6 figs., 8 tabs.

  4. Rod consolidation of RG and E's (Rochester Gas and Electric Corporation) spent PWR (pressurized water reactor) fuel

    SciTech Connect

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister.

  5. FUEL PERFORMANCE IMPROVEMENT PROGRAM Power-Ramp Testing and Postirradiation Examination of PCI- Resistant LWR Fuel Rod Designs

    SciTech Connect

    Barner, J. O.; Guenther, R. J.

    1982-09-01

    This report describes the power-ramp testing results from 10 fuel rods irradiated in the Halden Boiling Water Reactor (HBWR), Halden, Norway. Tne work is part of the Fuel Performance Improvement Program (FPIP), which is sponsored by the U.S. Department of Energy (DUE) and is conducted through the joint efforts of Consumers Power Company, Exxon Nuclear Company, lnc., and Pacific Northwest Laboratory. The objective of the FPlP is to identify and demonstrate fuel concepts with improved pellet-cladding interaction (PCl) behavior that will be capable of extended burnup. The postirradiation examination results obtained from one nonramped rod are also presented. The power-ramping behavior of three basic fuel rod types--rods with annular-pellet fuel, sphere-pac fuel, and dished-pellet (reference) fuel--are compared in terms of mechanisms known to promote PCl failures. The effects of graphite coating on the inside cladding surface and helium pressurization in rods witn annular fuel are also evaluated .

  6. Development and control of the process for the manufacture of zircaloy-4 tubing for LWBR fuel rods

    SciTech Connect

    Eyler, J.H.

    1981-01-01

    The technical requirements for the Light Water Breeder Reactor (LWBR) fuel elements (fuel rods) imposed certain unique requirements for the low hafnium Zircaloy-4 tubing used as fuel rod cladding. This report describes, in detail, the tube manufacturing process, the product and process controls used, the inspections and tests performed, and the efforts involved in refining a commercial tube reducing process to produce tubes that would satisfy the requirements for LWBR fuel rod cladding.

  7. 3D Simulation of Missing Pellet Surface Defects in Light Water Reactor Fuel Rods

    SciTech Connect

    B.W. Spencer; J.D. Hales; S.R. Novascone; R.L. Williamson

    2012-09-01

    The cladding on light water reactor (LWR) fuel rods provides a stable enclosure for fuel pellets and serves as a first barrier against fission product release. Consequently, it is important to design fuel to prevent cladding failure due to mechanical interactions with fuel pellets. Cladding stresses can be effectively limited by controlling power increase rates. However, it has been shown that local geometric irregularities caused by manufacturing defects known as missing pellet surfaces (MPS) in fuel pellets can lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. Nuclear fuel performance codes commonly use a 1.5D (axisymmetric, axially-stacked, one-dimensional radial) or 2D axisymmetric representation of the fuel rod. To study the effects of MPS defects, results from 1.5D or 2D fuel performance analyses are typically mapped to thermo-mechanical models that consist of a 2D plane-strain slice or a full 3D representation of the geometry of the pellet and clad in the region of the defect. The BISON fuel performance code developed at Idaho National Laboratory employs either a 2D axisymmetric or 3D representation of the full fuel rod. This allows for a computational model of the full fuel rod to include local defects. A 3D thermo-mechanical model is used to simulate the global fuel rod behavior, and includes effects on the thermal and mechanical behavior of the fuel due to accumulation of fission products, fission gas production and release, and the effects of fission gas accumulation on thermal conductivity across the fuel-clad gap. Local defects can be modeled simply by including them in the 3D fuel rod model, without the need for mapping between two separate models. This allows for the complete set of physics used in a fuel performance analysis to be included naturally in the computational representation of the local defect, and for the effects of the

  8. Fission gas release restrictor for breached fuel rod

    DOEpatents

    Kadambi, N. Prasad; Tilbrook, Roger W.; Spencer, Daniel R.; Schwallie, Ambrose L.

    1986-01-01

    In the event of a breach in the cladding of a rod in an operating liquid metal fast breeder reactor, the rapid release of high-pressure gas from the fission gas plenum may result in a gas blanketing of the breached rod and rods adjacent thereto which impairs the heat transfer to the liquid metal coolant. In order to control the release rate of fission gas in the event of a breached rod, the substantial portion of the conventional fission gas plenum is formed as a gas bottle means which includes a gas pervious means in a small portion thereof. During normal reactor operation, as the fission gas pressure gradually increases, the gas pressure interiorly of and exteriorly of the gas bottle means equalizes. In the event of a breach in the cladding, the gas pervious means in the gas bottle means constitutes a sufficient restriction to the rapid flow of gas therethrough that under maximum design pressure differential conditions, the fission gas flow through the breach will not significantly reduce the heat transfer from the affected rod and adjacent rods to the liquid metal heat transfer fluid flowing therebetween.

  9. Fuel integrity consequences of a misaligned control rod incident: Final report

    SciTech Connect

    Husser, D.L.; Delano, B.J.; Crist, S.H.; Mayer, J.T.; Lewis, L.Y.; Harris, K.L.

    1987-04-01

    During cycle 6 operation of the Arkansas Nuclear One Unit 1 reactor, an unanticipated transient occurred as a result of the rapid withdrawal at full power of a misaligned (27% withdrawn) control rod assembly (CRA). In less than one hour, operators realigned the assembly with the remaining rods in its group. Since the removal of the misaligned CRA was known to have caused high local power changes, the preliminary assessment was that stress corrosion cracking (SCC) occurred in the rods directly affected by the withdrawal. The potentially affected fuel assembly and certain selected additional assemblies were inspected using the Babcock and Wilcox ECHO-330 System, which permits the identification of individual rod failures. Based on the data gathered during this project, the misalignment event resulted in no occurrences of SCC-related fuel rod failures. The absence of failed rods in the assembly most significantly affected by the withdrawal clearly eliminates the SCC failure mode from consideration. The details of the power transient should be sufficient as benchmark cases to develop and verify computer codes designed to model power shock events in Zircaloy-clad fuel rods. The project is applicable to both SCC failure modeling and to such areas as load-following and power recovery operations where significant control rod movement is required. The power shock event meets the project objectives by providing a no-failure case under conditions approaching or exceeding the power change and levels typically associated with the SCC failure mode. The event also confirms the ability of pressurized water reactor fuel rods to sustain large power shocks without adverse effects.

  10. The COPERNIC3 project: how AREVA is successfully developing an advanced global fuel rod performance code

    SciTech Connect

    Garnier, Ch.; Mailhe, P.; Sontheimer, F.; Landskron, H.; Deuble, D.; Arimescu, V.I.; Billaux, M.

    2007-07-01

    Fuel performance is a key factor for minimizing operating costs in nuclear plants. One of the important aspects of fuel performance is fuel rod design, based upon reliable tools able to verify the safety of current fuel solutions, prevent potential issues in new core managements and guide the invention of tomorrow's fuels. AREVA is developing its future global fuel rod code COPERNIC3, which is able to calculate the thermal-mechanical behavior of advanced fuel rods in nuclear plants. Some of the best practices to achieve this goal are described, by reviewing the three pillars of a fuel rod code: the database, the modelling and the computer and numerical aspects. At first, the COPERNIC3 database content is described, accompanied by the tools developed to effectively exploit the data. Then is given an overview of the main modelling aspects, by emphasizing the thermal, fission gas release and mechanical sub-models. In the last part, numerical solutions are detailed in order to increase the computational performance of the code, with a presentation of software configuration management solutions. (authors)

  11. Remote Gamma Scanning System for Characterization of BWR and PWR Fuel Rod Sections

    SciTech Connect

    Crowell, Shannon L.; Alzheimer, James M.

    2011-08-08

    Sometimes challenges with the design and deployment of automated equipment in remote environments deals more with the constraints imposed by the remote environment than it does with the details of the automation. This paper discusses the development of a scanning system used to provide gamma radiation profiles of irradiated fuel rod segments. The system needed the capability to provide axial scans of cut segments of BWR and PWR fuel rods. The scanning location is A-Cell at the Radiochemical Processing Laboratory (RPL) at the Hanford site in Washington State. The criteria for the scanning equipment included axial scanning increments of a tenth of an inch or less, ability to scan fuel rods with diameters ranging from 3/8 inch to 5/8 inch in diameter, and fuel rod segments up to seven feet in length. Constraints imposed by the environment included having the gamma detector and operator controls on the outside of the hot cell and the scanning hardware on the inside of the hot cell. This entailed getting a narrow, collimated beam of radiation from the fuel rod to the detector on the outside of the hot cell while minimizing the radiation exposure caused by openings for the wires and cables traversing the hot cell walls. Setup and operation of all of the in-cell hardware needed to accommodate limited access ports and use of hot cell manipulators. The radiation levels inside the cell also imposed constraints on the materials used.

  12. Investigation of stainless steel clad fuel rod failures and fuel performance in the Connecticut Yankee Reactor. Final report

    SciTech Connect

    Pasupathi, V.; Klingensmith, R. W.

    1981-11-01

    Significant levels of fuel rod failures were observed in the batch 8 fuel assemblies of the Connecticut Yankee reactor. Failure of 304 stainless steel cladding in a PWR environment was not expected. Therefore a detailed poolside and hot cell examination program was conducted to determine the cause of failure and identify differences between batch 8 fuel and previous batches which had operated without failures. Hot cell work conducted consisted of detailed nondestructive and destructive examination of fuel rods from batches 7 and 8. The results indicate that the batch 8 failure mechanism was stress corrosion cracking initiating on the clad outer surface. The sources of cladding stresses are believed to be (a) fuel pellet chips wedged in the cladding gap, (b) swelling of highly nondensifying batch 8 fuel and (c) potentially harmful effects of a power change event that occurred near the end of the second cycle of irradiation for batch 8.

  13. Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

    SciTech Connect

    Ellis, Ronald James

    2015-04-01

    The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

  14. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    SciTech Connect

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-07-25

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on {approx}0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford ''New Production Reactor'', later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory's (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of uranium and ranged in

  15. Rod bundle thermal-hydraulic and melt progression analysis of CORA severe fuel damage experiments

    SciTech Connect

    Suh, K.Y. )

    1994-04-01

    An integral, fast-running computational model is developed to simulate the thermal-hydraulic and melt progression behavior in a nuclear reactor rod bundle under severe fuel damage conditions. This consists of the submodels for calculating steaming from the core, hydrogen formation, heat transfer in and out of the core, cooling from core spray or injection, and, most importantly, fuel melting, relocation, and freezing with chemical interactions taking place among the material constituents in a degrading core. The integral model is applied to three German severe fuel damage tests to analyze the core thermal and melt behavior: CORA-16 (18-rod bundle and slow cooling), CORA-17 (18-rod bundle and quenching), and CORA-18 (48-rod bundle and slow cooling). Results of the temperature response of the fuel rods, the channel box, and the absorber blade; hydrogen generation from the fuel rod and the channel box; and core material eutectic formation, melt relocation, and blockage formation are discussed. Reasonable agreement is observed for component temperatures at midelevation where prediction and measurement uncertainties are minimal. However, discrepancies or uncertainties are noticed for hydrogen generation and core-melt progression. The experimentally observed peak generation of hydrogen upon reflooding is not able to be reproduced, and the total amount generated is generally underpredicted primarily because of the early relocation of the Zircaloy fuel channel box and cladding. Also, difficulties are encountered in the process of assessing the core-melt formation and the relocation model because of either modeling uncertainties or a lack of definitive metallurgical data as a function of time throughout the transient.

  16. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    SciTech Connect

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.

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

    NASA Astrophysics Data System (ADS)

    Najeeb, Umair

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

  18. Fuel Rod Thermal-Mechanical Behavior, Versions FRAPCON2, FRAPCON2/VIM4 & FRAPCON2/VIM5.

    2002-03-25

    Version 02 This package contains three versions of the FRAPCON series of fuel rod response modeling programs. The FRAPCON series, like the earlier FRAP-S and GAPCON-THERMAL codes, is designed to predict the steady-state long-term burnup response of oxide fuel rods in light water reactors (LWRs). In addition, these codes generate the initial conditions for transient fuel rod analysis by the FRAP-T6 or thermal-hydraulic analysis programs. The FRAPCON2 programs calculate the temperature, pressure, deformation, and failuremore » histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include heat conduction through the fuel and cladding, cladding elastic and plastic deformation, fuel-cladding mechanical interaction, fission gas release, fuel rod internal gas pressure, heat transfer between fuel and cladding, cladding oxidation, and heat transfer from cladding to coolant. Material properties, water properties, and heat transfer correlation data are included. The FRAPCON series replaced the FRAP-S1, FRAP-S2, and FRAP-S3 series of programs. The fuel temperature computation used in the FRAPCON series was taken from the GAPCON-THERMAL2 code (NESC 618). FRAPCON2/VIM4 generates the initial conditions for transient fuel rod analysis used either by FRAP-T6 (NESC 658) or RELAP4/MOD7 (NESC 369).« less

  19. Preliminary design and manufacturing feasibility study for a machined Zircaloy triangular pitch fuel rod support system (grids) (AWBA development program)

    SciTech Connect

    Horwood, W A

    1981-07-01

    General design features and manufacturing operations for a high precision machined Zircaloy fuel rod support grid intended for use in advanced light water prebreeder or breeder reactor designs are described. The grid system consists of a Zircaloy main body with fuel rod and guide tube cells machined using wire EDM, a separate AM-350 stainless steel insert spring which fits into a full length T-slot in each fuel rod cell, and a thin (0.025'' or 0.040'' thick) wire EDM machined Zircaloy coverplate laser welded to each side of the grid body to retain the insert springs. The fuel rods are placed in a triangular pitch array with a tight rod-to-rod spacing of 0.063 inch nominal. Two dimples are positioned at the mid-thickness of the grid (single level) with a 90/sup 0/ included angle. Data is provided on the effectiveness of the manufacturing operations chosen for grid machining and assembly.

  20. Finite-element procedure for calculating the three-dimensional inelastic bowing of fuel rods (AWBA development program)

    SciTech Connect

    Martin, S E

    1982-05-01

    An incremental finite element procedure is developed for calculating the in-pile lateral bowing of nuclear fuel rods. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the results of an axisymmetric stress analysis of the fuel rod. The effects which are taken into account in calculating the rod's lateral bowing include: (a) lateral, axial, and rotational motions and forces at the rod supports, (b) transverse gradients of temperature, fast-neutron flux, and fissioning rate, and (c) cladding circumferential wall thickness variation. The procedure developed in this report could be used to form the basis for a computer program to calculate the time-dependent bowing as a function of the fuel rod's operational and environmental history.

  1. Investigation of Minimum Film boiling Phenomena on Fuel Rods Under Blowdown Cooling Conditions

    SciTech Connect

    Stephen M. Bajorek; Michael Gawron; Timothy Etzel; Lucas Peterson

    2003-06-30

    Blowdon cooling heat transfer is an important process that occurs early in a hypothetical large break loss-of-coolant accident (LOCA) in a pressurized water reactor. During blowdown, the flow through the hot assembly is a post-critical heat flux dispersed droplet flow. The heat transfer mechanisms that occur in blowdown cooling are complex and depend on droplet and heated surface interaction. In a safety analysis, it is of considerable importance to determine the thermal-hydraulic conditions leading to the minimum film boiling temperature, Tmin. A flow boiling rig for measurement of blowdown cooling heat transfer and quench phenomena on a nuclear fuel rod simulator was designed and constructed for operation at up to 12.4 MPa. The test section consisted of a concentric annulus, with a 9.5 mm OD nuclear fuel rod simulator at the center. The rod was contained within a 0.85 mm thick, 19 mm OD 316 stainless steel tube, forming the flow channel. Two types of rods were tested; one type was sheathed with Inconel 600 while the other was clad with Zircaloy-2. Water was injected into the test section at the top of the heated length through an injection header. This header was an annular sign that fit around the fuel rod simulator and within the stainless steel tube. Small spacers aligned the injection header and prevented contract with either the heater rod or the tube. A series of small diameter holes at the bottom of the header caused the formation of droplets that became entrained with the steam flow. The test section design was such that quench would take place on the rod, and not along the channel outer annulus.

  2. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    SciTech Connect

    Kee, E.

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  3. Development of techniques for joining fuel rod simulators to test assemblies

    SciTech Connect

    Moorhead, A.J.; Reed, R.W.

    1980-01-01

    A unique tubular electrode carrier is described for gas tungsten-arc welding small-diameter nuclear fuel rod simulators to the tubesheet of a test assembly. Both the close-packed geometry of the array of simulators and the extension of coaxial electrical conductors from each simulator hindered access to the weld joint. Consequently, a conventional gas tungsten-arc torch could not be used. Two seven-rod assemblies that were mockups of the simulator-to-tubesheet joint area were welded and successfully tested. Modified versions of the electrode carrier for brazing electrical leads to the upper ends of the fuel pin simulators are also described. Satisfactory brazes have been made on both single-rod mockups and an array of 25 simulators by using the modified electrode carrier and a filler metal with a composition of 71.5 Ag-28 Cu-0.5 Ni.

  4. Experimental investigation of laboratory-scale rocket engine fed on solid polyethylene rod as fuel

    NASA Astrophysics Data System (ADS)

    Yemets, V. V.; Sanin, F. P. Dzhur, Ye. O.; Masliany, M. V.; Kostritsyn, O. Yu.; Minteev, G. V.; Ushkanov, V. M.

    Fire testing of the laboratory-scale rocket engine with the consumable solid polyethylene rod as fuel is described. The experimental data on heat flows, gasification rate and heat transfer coefficient are presented. Results of the testing may be useful for designing launch vehicles with combustible polyethylene tank shells.

  5. DESTRUCTIVE EXAMINATION OF 3-CYCLE LWR (LIGHT WATER REACTOR) FUEL RODS FROM TURKEY POINT UNIT 3 FOR THE CLIMAX - SPENT FUEL TEST

    SciTech Connect

    ATKIN SD

    1981-06-01

    The destructive examination results of five light water reactor rods from the Turkey Point Unit 3 reactor are presented. The examinations included fission gas collection and analyses, burnup and hydrogen analyses, and a metallographic evaluation of the fuel, cladding, oxide, and hydrides. The rods exhibited a low fission gas release with all other results appearing representative for pressurized water reator fuel rods with similar burnups (28 GWd/MTU) and operating histories.

  6. Dealing with Historical Discrepancies: The Recovery of National Research Experiment (NRX) Reactor Fuel Rods at Chalk River Laboratories (CRL) - 13324

    SciTech Connect

    Vickerd, Meggan

    2013-07-01

    Following the 1952 National Research Experiment (NRX) Reactor accident, fuel rods which had short irradiation histories were 'temporarily' buried in wooden boxes at the 'disposal grounds' during the cleanup effort. The Nuclear Legacy Liabilities Program (NLLP), funded by Natural Resources Canada (NRCan), strategically retrieves legacy waste and restores lands affected by Atomic Energy of Canada Limited (AECL) early operations. Thus under this program the recovery of still buried NRX reactor fuel rods and their relocation to modern fuel storage was identified as a priority. A suspect inventory of NRX fuels was compiled from historical records and various research activities. Site characterization in 2005 verified the physical location of the fuel rods and determined the wooden boxes they were buried in had degraded such that the fuel rods were in direct contact with the soil. The fuel rods were recovered and transferred to a modern fuel storage facility in 2007. Recovered identification tags and measured radiation fields were used to identify the inventory of these fuels. During the retrieval activity, a discrepancy was discovered between the anticipated number of fuel rods and the number found during the retrieval. A total of 32 fuel rods and cans of cut end pieces were recovered from the specified site, which was greater than the anticipated 19 fuel rods and cans. This discovery delayed the completion of the project, increased the associated costs, and required more than anticipated storage space in the modern fuel storage facility. A number of lessons learned were identified following completion of this project, the most significant of which was the potential for discrepancies within the historical records. Historical discrepancies are more likely to be resolved by comprehensive historical record searches and site characterizations. It was also recommended that a complete review of the wastes generated, and the total affected lands as a result of this historic

  7. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    SciTech Connect

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-09-15

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on ~0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford “New Production Reactor,” later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory’s (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of uranium and ranged in Plutonium

  8. Models for the Configuration and Integrity of Partially Oxidized Fuel Rod Cladding at High Temperatures

    SciTech Connect

    Siefken, L.J.

    1999-01-01

    Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. These models are expected to improve the calculations of several important aspects of fuel rod behavior. First, an improved mapping was established from a compilation of PIE results from severe fuel damage tests of the configuration of melted metallic cladding that is retained by an oxide layer. The improved mapping accounts for the relocation of melted cladding in the circumferential direction. Then, rules based on PIE results were established for calculating the effect of cladding that has relocated from above on the oxidation and integrity of the lower intact cladding upon which it solidifies. Next, three different methods were identified for calculating the extent of dissolution of the oxidic part of the cladding due to its contact with the metallic part. The extent of dissolution effects the stress and thus the integrity of the oxidic part of the cladding. Then, an empirical equation was presented for calculating the stress in the oxidic part of the cladding and evaluating its integrity based on this calculated stress. This empirical equation replaces the current criterion for loss of integrity which is based on temperature and extent of oxidation. Finally, a new rule based on theoretical and experimental results was established for identifying the regions of a fuel rod with oxidation of both the inside and outside surfaces of the cladding. The implementation of these models is expected to eliminate the tendency of the SCDAP/RELAP5 code to overpredict the extent of oxidation of the upper part of fuel rods and to underpredict the extent of oxidation of the lower part of fuel rods and the part with a high concentration of relocated material. This report is a revision and reissue of the report entitled, Improvements in Modeling of Cladding Oxidation and Meltdown.

  9. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    SciTech Connect

    Chung, H.M. )

    1989-09-01

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 {mu}m in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307{degree}C rather than the normal 288{degree}C, a relatively thick (50 to 70 {mu}m) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs.

  10. On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

    PubMed

    Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

    2016-08-01

    For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line. PMID:27209090

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

    NASA Astrophysics Data System (ADS)

    Williamson, R. L.

    2011-08-01

    A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO 2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

  12. Enhancing the ABAQUS Thermomechanics Code to Simulate Steady and Transient Fuel Rod Behavior

    SciTech Connect

    R. L. Williamson; D. A. Knoll

    2009-09-01

    A powerful multidimensional fuels performance capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth , gap heat transfer, and gap/plenum gas behavior during irradiation. The various modeling capabilities are demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multi-pellet fuel rod, during both steady and transient operation. Computational results demonstrate the importance of a multidimensional fully-coupled thermomechanics treatment. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermo-mechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

  13. External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the LOFT reactor

    SciTech Connect

    Welty, R. K.

    1980-01-01

    The Exxon Nuclear Company, Inc., acting as a Subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications, and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis.

  14. Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences

    SciTech Connect

    Bailey, W.J.

    1990-02-01

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs.

  15. Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents

    SciTech Connect

    Siefken, Larry James

    1999-02-01

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the clad-ding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; "Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents."

  16. Observation of Porosity Reduction in a Densification-Prone Test Fuel Rod: Data and Analysis

    SciTech Connect

    Cunningham, M. E.; Daniel, J. L.; Lanning, D. D.

    1981-10-01

    Instrumented fuel assembly (IFA)-431 was irradiated in the Halden Boiling Water Reactor (HBWR) for the purpose of extending the steady-state data base. Rod 6 of this assembly began irradiation with UO{sub 2} fuel of 92% theoretical density (TD) that was unstable with respect to in-reactor densification. Thermal resintering tests resulted in a final density of 95.3% TD while post-irradiation examination (PIE) indicated a final density of 96.5% TD. Observed microstructural changes were consistent with published densification studies; there was a marked depletion of submicrometer diameter pores and total pore volume. However, grain size increased only slightly, indicating that internal pellet temperatures did not reach the 1875K applied in resintering tests. Oensification was observed to increase the temperatures in rod 6, but temperatures did not become as high as for a sibling rod that simulated instantaneous densification. Temperatures calculated with U.S. Nuclear Regulatory Commission (NRC) fuel performance computer codes were generally higher than observed temperatures.

  17. A New Insight into Energy Distribution of Electrons in Fuel-Rod Gap in VVER-1000 Nuclear Reactor

    NASA Astrophysics Data System (ADS)

    Fereshteh, Golian; Ali, Pazirandeh; Saeed, Mohammadi

    2015-06-01

    In order to calculate the electron energy distribution in the fuel rod gap of a VVER-1000 nuclear reactor, the Fokker-Planck equation (FPE) governing the non-equilibrium behavior of electrons passing through the fuel-rod gap as an absorber has been solved in this paper. Besides, the Monte Carlo Geant4 code was employed to simulate the electron migration in the fuel-rod gap and the energy distribution of electrons was found. As for the results, the accuracy of the FPE was compared to the Geant4 code outcomes and a satisfactory agreement was found. Also, different percentage of the volatile and noble gas fission fragments produced in fission reactions in fuel rod, i.e. Krypton, Xenon, Iodine, Bromine, Rubidium and Cesium were employed so as to investigate their effects on the electrons' energy distribution. The present results show that most of the electrons in the fuel rod's gap were within the thermal energy limitation and the tail of the electron energy distribution was far from a Maxwellian distribution. The interesting outcome was that the electron energy distribution is slightly increased due to the accumulation of fission fragments in the gap. It should be noted that solving the FPE for the energy straggling electrons that are penetrating into the fuel-rod gap in the VVER-1000 nuclear reactor has been carried out for the first time using an analytical approach.

  18. Development of the ACP safeguards neutron counter for PWR spent fuel rods

    NASA Astrophysics Data System (ADS)

    Lee, Tae-Hoon; Menlove, Howard O.; Lee, Sang-Yoon; Kim, Ho-Dong

    2008-04-01

    An advanced neutron multiplicity counter has been developed for measuring spent fuel in the Advanced spent fuel Conditioning Process (ACP) at the Korea Atomic Energy Research Institute (KAERI). The counter uses passive neutron multiplicity counting to measure the 244Cm content in spent fuel. The input to the ACP process is spent fuel from pressurized water reactors (PWRs), and the high intensity of the gamma-ray exposure from spent fuel requires a careful design of the counter to measure the neutrons without gamma-ray interference. The nuclear safeguards for the ACP facility requires the measurement of the spent fuel input to the process and the Cm/Pu ratio for the plutonium mass accounting. This paper describes the first neutron counter that has been used to measure the neutron multiplicity distribution from spent fuel rods. Using multiple samples of PWR spent fuel rod-cuts, the singles (S), doubles (D), and triples (T) rates of the neutron distribution for the 244Cm nuclide were measured and calibration curves were produced. MCNPX code simulations were also performed to obtain the three counting rates and to compare them with the measurement results. The neutron source term was evaluated by using the ORIGEN-ARP code. The results showed systematic difference of 21-24% in the calibration graphs between the measured and simulation results. A possible source of the difference is that the burnup codes have a 244Cm uncertainty greater than ±15% and it would be systematic for all of the calibration samples. The S/D and D/T ratios are almost constant with an increment of the 244Cm mass, and this indicates that the bias is in the 244Cm neutron source calculation using the ORIGEN-ARP source code. The graphs of S/D and D/T ratios show excellent agreement between measurement and MCNPX simulation results.

  19. FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

    SciTech Connect

    Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

    1981-01-01

    FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

  20. Scaled-up dual anode/cathode microbial fuel cell stack for actual ethanolamine wastewater treatment.

    PubMed

    An, Byung-Min; Heo, Yoon; Maitlo, Hubdar-Ali; Park, Joo-Yang

    2016-06-01

    The aim of this work was to develop the scale-up microbial fuel cell technology for actual ethanolamine wastewater treatment, dual anode/cathode MFC stacks connected in series to achieve any desired current, treatment capacity, and volume capacity. However, after feeding actual wastewater into the MFC, maximum power density decreased while the corresponding internal resistance increased. With continuous electricity production, a stack of eight MFCs in series achieved 96.05% of COD removal and 97.30% of ammonia removal at a flow rate of 15.98L/d (HRT 12h). The scaled-up dual anode/cathode MFC stack system in this research was demonstrated to treat actual ETA wastewater with the added benefit of harvesting electricity energy. PMID:26888335

  1. Test-fuel power-coupling dependence on TREAT control-rod positions

    SciTech Connect

    Harrison, L.J.; Klotzkin, G.; Hart, P.R.; Swanson, R.W.

    1983-01-01

    The Transient Reactor Test (TREAT) is a graphite moderated, UO/sub 2/ fueled test reactor located at the Idaho National Engineering Laboratory and operated by Argonne National Laboratory. Test fuel is placed in containment vessels in the center of the reactor and subjected to computer-controlled transient irradiations which can result in experimental fuel melting or even vaporizing. The reactor was designed to have a strong negative temperature coefficient and to operate adiabatically. Consequently large reactivity insertions, up to 6.2% ..delta..k/k, may be required during a transient as the core temperature increases as much as 570/sup 0/C. This reactivity insertion is accomplished typically over 10 to 20 seconds by hydraulically actuated transient control rods. Evaluation of empirical data has indicated that control-rod-position changes cause power-coupling changes during a transient and usually are the primary factor in determining the ratio of the transient-averaged to steady-state test-fuel power coupling.

  2. Experimental Study on the Influence of the Supporting Condition and Rod Motion on the Fuel Fretting Damage

    SciTech Connect

    Kim, Hyung-Kyu; Lee, Young-Ho

    2007-07-01

    Present study focuses on the influence of a supporting condition and a rod motion on a fuel fretting wear through experiments using a self-developed wear simulator, which was presented at the Water Reactor Fuel Performance Meeting, Kyoto Japan in 2005. In the experiment, a fuel rod specimen of two span lengths is vibrated by two perpendicularly aligned electromagnetic actuators. Both ends of the rod specimen are supported with a positive contact force and a variation of the supporting condition is simulated by moving each of the four grid strap specimens constituting a center grid cell. As for the supporting condition, 0.1 mm gap and 10 N force are used; a circular and a diagonal traces are applied for the rod motion. The contact shape of the spring/dimple is concave, to try and increase the contact area. Both the spring/dimple and fuel rod specimens were fabricated from the as-received materials (zirconium alloy) for a commercial fuel assembly. Experiments are carried out under a room temperature and distilled water condition. Experiment of each condition is carried out for 72 hours. Wear volume, area and depth on the cladding tubes are examined. As a result, the present concave shaped spring/dimple causes less wear when the rod moves in a circular manner than a diagonal one if there is a positive contact force (10 N). However, a diagonal motion causes more wear when a gap (0.1 mm) exists. Wear amount at the spring and dimple is influenced by the location of them and the rod motion. It is found that the wear is concentrated at the contact edges between the spring/dimple and rods due to the contact shape. The influence of the rod motion on the worn area and its shape is also discussed. (authors)

  3. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Jiang, Hao

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  4. Measurement of Fresh Fuel Rods to Demonstrate Compliance with Criticality Safety Limits

    SciTech Connect

    Miko, David K.; Desimone, David J.

    2015-11-03

    In order to operate TA-66 as a radiological facility with the quantity of nuclear material required to fulfil its mission, a criticality safety evaluation was required. This evaluation defined the control parameters for operations at the facility. The resulting evaluation for TA-66 placed limits on the amount of SNM, as well as other materials such as beryllium. In addition, there is a limit on the number of uranium fuel rods allowed subject to enrichment, outer diameter, and overall length restrictions. The enrichments for the rods to be shipped to TA-66 were documented in LA-UR-13-23581, but the outer diameter and length were not documented. This report provides this information.

  5. Fuel Rod Melt Progression Simulation Using Low-Temperature Melting Metal Alloy

    SciTech Connect

    Seung Dong Lee; Suh, Kune Y.; GoonCherl Park; Un Chul Lee

    2002-07-01

    The TMI-2 accident and various severe fuel damage experiments have shown that core damage is likely to proceed through various states before the core slumps into the lower head. Numerous experiments were conducted to address when and how the core can lose its original geometry, what geometries are formed, and in what processes the core materials are transported to the lower plenum of the reactor pressure vessel. Core degradation progresses along the line of clad ballooning, clad oxidation, material interaction, metallic blockage, molten pool formation, melt progression, and relocation to the lower head. Relocation into the lower plenum may occur from the lateral periphery or from the bottom of the core depending upon the thermal and physical states of the pool. Determining the quantities and rate of molten material transfer to the lower head is important since significant amounts of molten material relocated to the lower head can threaten the vessel integrity by steam explosion and thermal and mechanical attack of the melt. In this paper the focus is placed on the melt flow regime on a cylindrical fuel rod utilizing the LAMDA (Lumped Analysis of Melting in Degrading Assemblies) facility at the Seoul National University. The downward relocation of the molten material is a combination of the external film flow and the internal pipe flow. The heater rods are 0.8 m long and are coated by a low-temperature melting metal alloy. The electrical internal heating method is employed during the test. External heating is adopted to simulate the exothermic Zircaloy-steam reaction. Tests are conducted in several quasi-steady-state conditions. Given the variable boundary conditions including the heat flux and the water level, observation is made for the melting location, progression, and the mass of molten material. Finally, the core melt progression model is developed from the visual inspection and quantitative analysis of the experimental data. As the core material relocates

  6. A cone beam computed tomography inspection method for fuel rod cladding tubes

    NASA Astrophysics Data System (ADS)

    Fu, Jian; Tan, Renbo; Wang, Qianli; Deng, Jingshan; Liu, Ming

    2012-10-01

    Fuel rods in nuclear power plants consist of UO2 pellets enclosed in Zirconium alloy (Zircaloy) cladding tube, which is composed of a body and a plug. The body is manufactured separately from the plug and, before its use, the plug is welded with the body. It is vitally important for the welding zone to remain free from defects after the fuel pellets are loaded into the cladding tube to prevent the radioactive fission products from leaking. X-ray computed tomography (CT) is in principle a feasible inspection method for the welding zone, but it faces several challenges due to the high attenuation of Zircaloy. In this paper, a cone beam CT method is proposed to address these issues and perform the welding flaw inspection. A Zircaloy compensator is adopted to narrow the signal range, a structure-based background removal technique to reveal the defects, a linear extension technique to determine the reference X-ray intensity signal and FDK algorithm to reconstruct the slice images. A prototype system, based on X-ray tube source and flat panel detector, has been developed and the experiments in this system have demonstrated that the welding void and the incomplete joint penetrations could be detected by this method. This approach may find applications in the quality control of nuclear fuel rods.

  7. Preliminary design report for modeling of hydrogen uptake in fuel rod cladding during severe accidents

    SciTech Connect

    Siefken, L.J.

    1998-08-01

    Preliminary designs are described for models of the interaction of Zircaloy and hydrogen and the consequences of this interaction on the behavior of fuel rod cladding during severe accidents. The modeling of this interaction and its consequences involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer at the cladding external surface, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental and theoretical results are presented that show the uptake of hydrogen in the event of dissolution of the oxide layer occurs rapidly and that show the release of hydrogen in the event of cracking of the cladding occurs rapidly. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert`s law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for Zr-H interaction into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the Zr-H interaction models on the calculated behavior of fuel rods in severe accident conditions.

  8. Analysis of MERCI decay heat measurement for PWR UO{sub 2} fuel rod

    SciTech Connect

    Jaboulay, J.C.; Bourganel, S.

    2012-01-15

    Decay heat measurements, called the MERCI experiment, were conducted at Commissariat a l'Energie Atomique (CEA)/Saclay to characterize accurately residual power at short cooling time and verify its prediction by decay code and nuclear data. The MOSAIC calorimeter, developed and patented by CEA/Grenoble (DTN/SE2T), enables measurement of the decay heat released by a pressurized water reactor (PWR) fuel rod sample between 200 and 4 W within a precision of 1%. The MERCI experiment included three phases. At first, a UO{sub 2} fuel rod sample was irradiated in the CEA/Saclay experimental reactor OSIRIS. The burnup achieved at the end of irradiation was similar to 3.5 GWd/tonne. The second phase was the transfer of the fuel rod sample from its irradiation location to a hot cell, to be inserted inside the MOSAIC calorimeter. It took 26 min to carry out the transfer. Finally, decay heat released by the PWR sample was measured from 27 min to 42 days after shutdown. Post irradiation examinations were performed to measure concentrations of some heavy nuclei (U, Pu) and fission products (Cs, Nd). The decay heat was predicted using a calculation scheme based on the PEPIN2 depletion code, the TRIPOLI-4 Monte Carlo code, and the JEFF3.1.1 nuclear data file. The MERCI experiment analysis shows that the discrepancy between the calculated and the experimental decay heat values is included between -10% at 27 min and +6% at 12 h, 30 min otter shutdown. From 4 up to 42 days of cooling time, the difference between calculation and measurement is about ± 1%, i.e., experimental uncertainty. The MERCI experiment represents a significant contribution for code validation; the time range above 10{sup 5} s has not been validated previously. (authors)

  9. Fuel rod and core materials investigations related to LWR extended burnup operation

    NASA Astrophysics Data System (ADS)

    Kolstad, Erik; Vitanza, Carlo

    1992-06-01

    The paper deals with tests and recent measurements related to extended burnup fuel performance and describes test facilities and results in the areas of waterside cladding corrosion and irradiation-assisted stress corrosion cracking (IASCC). Fuel temperature data suggest a gradual degradation of UO 2 thermal conductivity with exposure in the range 6-8% per 10 MWd/kgUO 2 at temperatures below 700°C. The effect on the fuel microstructure of interlinkage and resintering phenomena is shown by measuring the surface-to-volume ( S/ V) ratio of the fuel. Changes in S/V with burnup are correlated to power rating and fuel operating temperature. No evidence was found of enhanced fission gas release during load-follow operation in the burnup range 25-45 MWd/kgUO 2. The effect of high lithium concentration (high pH) on the corrosion behaviour of pre-irradiated high burnup Zircaloy-4 fuel rods subjected either to nucleate boiling or to one-phase cooling conditions was studied. The oxide thickness growth rates measured at an average burnup up to 40 MWd/kgUO 2 are consistent with literature data and show no evidence of corrosion enhancement due to the high lithium content and little effect of cooling regime. A test facility for exploring the effects of environmental variables on IASCC behaviour of in-core structural materials is described.

  10. A Statistical Approach to Predict the Failure Enthalpy and Reliability of Irradiated PWR Fuel Rods During Reactivity-Initiated Accidents

    SciTech Connect

    Nam, Cheol; Jeong, Yong-Hwan; Jung, Youn-Ho

    2001-11-15

    During the last decade, the failure behavior of high-burnup fuel rods under a reactivity-initiated accident (RIA) condition has been a serious concern since fuel rod failures at low enthalpy have been observed. This has resulted in the reassessment of existing licensing criteria and failure-mode study. To address the issue, a statistics-based methodology is suggested to predict failure probability of irradiated fuel rods under an RIA. Based on RIA simulation results in the literature, a failure enthalpy correlation for an irradiated fuel rod is constructed as a function of oxide thickness, fuel burnup, and pulse width. Using the failure enthalpy correlation, a new concept of ''equivalent enthalpy'' is introduced to reflect the effects of the three primary factors as well as peak fuel enthalpy into a single damage parameter. Moreover, the failure distribution function with equivalent enthalpy is derived, applying a two-parameter Weibull statistical model. Finally, the sensitivity analysis is carried out to estimate the effects of burnup, corrosion, peak fuel enthalpy, pulse width, and cladding materials used.

  11. Synthesis and Analysis of Alpha Silicon Carbide Components for Encapsulation of Fuel Rods and Pellets

    SciTech Connect

    Kevin M. McHugh; John E. Garnier; George W. Griffith

    2011-09-01

    The chemical, mechanical and thermal properties of silicon carbide (SiC) along with its low neutron activation and stability in a radiation field make it an attractive material for encapsulating fuel rods and fuel pellets. The alpha phase (6H) is particularly stable. Unfortunately, it requires very high temperature processing and is not readily available in fibers or near-net shapes. This paper describes an investigation to fabricate a-SiC as thin films, fibers and near-net-shape products by direct conversion of carbon using silicon monoxide vapor at temperatures less than 1700 C. In addition, experiments to nucleate the alpha phase during pyrolysis of polysilazane, are also described. Structure and composition were characterized using scanning electron microscopy, energy dispersive spectroscopy and X-ray diffraction. Preliminary tensile property analysis of fibers was also performed.

  12. Experimental and numerical study on lead-bismuth heat transfer in a fuel rod simulator

    NASA Astrophysics Data System (ADS)

    Ma, Weimin; Karbojian, Aram; Hollands, Thorsten; Koch, Marco K.

    2011-08-01

    As a task of the EU project IP EUROTRANS towards development of an Accelerator Driven System (ADS) dedicated to the transmutation of long-lived fission products, experiments and simulations were performed on the TALL test facility at KTH to investigate thermal hydraulics along a single fuel rod simulator cooled by lead-bismuth eutectic (LBE). The fuel rod simulator is concentrically inserted in a tube, so that an annular channel is formed for LBE flow. This paper presents the measured temperature profiles in the annular channel, and the comparisons with the simulation results of the CFX code. The primary objective is to help understanding the LBE heat transfer characteristics and qualifying the turbulence and heat transfer modeling for LBE application. The quantitative comparison between the calculated and measured temperatures of the LBE indicates that the simulation underestimates the experiment at most radial and axial positions. Finally the uncertainties in measurement and the deficiency in turbulence models resulting in such a disagreement were discussed, which will be directive and beneficial to future work in the field.

  13. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    DOEpatents

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  14. High temperature postirradiation materials performance of spent pressurized water reactor fuel rods under dry storage conditions

    SciTech Connect

    Einziger, R.E.; Atkin, S.D.; Pasupathi, V.; Stellrecht, D.E.

    1982-04-01

    Postirradiation studies on failure mechanisms of well-characterized pressurized water reactor rods were conducted for up to a year at 482, 510, and 571/sup 0/C in limited air and inert gas atmospheres. No cladding breaches occurred even though the tests operated many orders of magnitude longer in time than the lifetime predicted by Blackburn's analyses. The extended lifetime is due to significant creep strain of the Zircaloy cladding, which decreases the internal rod pressure. The cladding creep also contributes to radial cracks, through the external oxide and internal fuel-cladding chemical interaction layers, which propagated into and arrested in an oxygen stabilized alpha-Zircaloy layer. There were no signs of either additional cladding hydriding, stress corrosion cracking, or fuel pellet degradation. If irradiation hardening does not reduce the stress rupture properties of Zircaloy, a conservative maximum storage temperature of 400/sup 0/C based on a stress-rupture mechanism is recommended to ensure a 1000-yr cladding lifetime.

  15. Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods

    SciTech Connect

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

  16. Thermo-Mechanical Analysis of Coated Particle Fuel Experiencing a Fast Control Rod Ejection Transient

    SciTech Connect

    Ortensi, J.; Brian Boer; Abderrafi M. Ougouag

    2010-10-01

    A rapid increase of the temperature and the mechanical stress is expected in TRISO coated particle fuel that experiences a fast Total Control Rod Ejection (CRE) transient event. During this event the reactor power in the pebble bed core increases significantly for a short time interval. The power is deposited instantly and locally in the fuel kernel. This could result in a rapid increase of the pressure in the buffer layer of the coated fuel particle and, consequently, in an increase of the coating stresses. These stresses determine the mechanical failure probability of the coatings, which serve as the containment of radioactive fission products in the Pebble Bed Reactor (PBR). A new calculation procedure has been implemented at the Idaho National Laboratory (INL), which analyzes the transient fuel performance behavior of TRISO fuel particles in PBRs. This early capability can easily be extended to prismatic designs, given the availability of neutronic and thermal-fluid solvers. The full-core coupled neutronic and thermal-fluid analysis has been modeled with CYNOD-THERMIX. The temperature fields for the fuel kernel and the particle coatings, as well as the gas pressures in the buffer layer, are calculated with the THETRIS module explicitly during the transient calculation. Results from this module are part of the feedback loop within the neutronic-thermal fluid iterations performed for each time step. The temperature and internal pressure values for each pebble type in each region of the core are then input to the PArticle STress Analysis (PASTA) code, which determines the particle coating stresses and the fraction of failed particles. This paper presents an investigation of a Total Control Rod Ejection (TCRE) incident in the 400 MWth Pebble Bed Modular reactor design using the above described calculation procedure. The transient corresponds to a reactivity insertion of $3 (~2000 pcm) reaching 35 times the nominal power in 0.5 seconds. For each position in the core

  17. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    SciTech Connect

    Johnson, G.L.

    1988-09-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degree}C and whether the cladding of the stored spent fuel ever exceeds 350{degree}C. Limiting the borehole to temperatures of 97{degree}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degree}C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degree}C for the full 1000-yr analysis period.

  18. Fuel performance improvement program: description and characterization of HBWR Series H-2, H-3, and H-4 test rods

    SciTech Connect

    Guenther, R.J.; Barner, J.O.; Welty, R.K.

    1980-03-01

    The fabrication process and as-built characteristics of the HBWR Series H-2 and H-3 test rods, as well as the three packed-particle (sphere-pac) rods in HBWR Series H-4 are described. The HBWR Series H-2, H-3, and H-4 tests are part of the irradiation test program of the Fuel Performance Improvement Program. Fifteen rods were fabricated for the three test series. Rod designs include: (1) a reference dished pellet design incorporating chamfered edges, (2) a chamfered, annular pellet design combined with graphite-coated cladding, and (3) a sphere-pac design. Both the annular-coated and sphere-pac designs include internal pressurization using helium.

  19. Issues in Three-Dimensional Depletion Analysis of Measured Data Near the End of a Fuel Rod

    SciTech Connect

    DeHart, Mark D; Gauld, Ian C; Suyama, Kenya

    2008-01-01

    The dynamics of reactor operation result in nonuniform axial-burnup profiles in fuel with any significant burnup. At the beginning of life in a pressurized water reactor (PWR), a near-cosine axial-shaped flux will begin depleting fuel near the axial center of a fuel assembly at a greater rate than at the ends. As the reactor continues to operate, the cosine flux shape will flatten because of the fuel depletion and fission-product buildup that occur near the center. However, because of the high leakage near the end of the fuel assembly, burnup will drop off rapidly near the ends. Partial-length absorbers or nonuniform axial fuel loadings can further complicate the burnup profile. In a boiling water reactor, the same phenomena come into play, but the burnup profile is complicated by the significant variation of axial moderator density and by nonuniform axial loadings of burnable poison rods. Numerous studies of axial burnup effects have been published. However, most analyses performed in estimation of isotopic distributions due to axial burnup have been based on a set of two-dimensional (2-D) calculations performed for burnups that represent the axial burnup distribution in a fuel assembly. In general, this approach works quite well because the in-core axial gradient of the neutron flux is small over most of the length of the fuel rod, and the 2-D approximation is appropriate. Conversely, because the axial gradient becomes significant as one approaches either end of the fuel assembly, the 2-D approximation begins to break down at that point. It has been theorized that axial leakage will lead to a reduced fast flux relative to the thermal flux, softening the spectrum near the ends of the fuel, and that a 2-D approximation is conservative in that it provides more plutonium production. This has not been put the test, however, for two reasons--a lack of good three-dimensional (3-D) analysis methods acceptable for away-from-reactor applications and, more importantly, a

  20. Rod examination gauge

    SciTech Connect

    Bacvinskas, W.S.; Bayer, J.E.; Davis, W.W.; Fodor, G.; Kikta, T.J.; Matchett, R.L.; Nilsen, R.J.; Wilczynski, R.

    1991-12-31

    The present invention is directed to a semi-automatic rod examination gauge for performing a large number of exacting measurements on radioactive fuel rods. The rod examination gauge performs various measurements underwater with remote controlled machinery of high reliability. The rod examination gauge includes instruments and a closed circuit television camera for measuring fuel rod length, free hanging bow measurement, diameter measurement, oxide thickness measurement, cladding defect examination, rod ovality measurement, wear mark depth and volume measurement, as well as visual examination. A control system is provided including a programmable logic controller and a computer for providing a programmed sequence of operations for the rod examination and collection of data.

  1. Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR

    SciTech Connect

    Yackle, T.R.; MacDonald, P.E.; Broughton, J.M.

    1980-01-01

    An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.

  2. Measurements of fuel pin/water hole worths and power peaking, void coefficients, and temperature coefficients for 4. 81 wt% enriched UO[sub 2] fuel rods

    SciTech Connect

    Harris, D.R.; Rohr, R.R.; Angelo, P.L.; Patrou, N.T.; Buckwheat, K.W.; Hayes, D.K.

    1990-01-01

    The Rensselaer Polytechnic Institute reactor critical facility is currently the only facility in North America providing critical measurement data in support of the light water reactor electric power industry. The reactor is fueled by 4.81 wt% [sup 235]U enriched UO[sub 2] high-density pellets in stainless steel clad fuel rods at the present time, although experiments with other fuels are being analyzed. The fuel pins are supported by inexpensive stainless steel lattice plates in a large open water tank. Three sets of lattice plates have been fabricated for fuel pins in square array with pitches 0.585, 0.613, and 0.640 in. (1.486, 0.613, and 1.656 cm, respectively) to provide a relevant range of water-to-fuel volume ratios. The measurements reported here are for the first of these, a relatively tight lattice of considerable interest for reactor physics methods for advanced fuels and reactors.

  3. Characterization of a suspect nuclear fuel rod in a case of illegal international traffic of fissile material.

    PubMed

    Capannesi, G; Vicini, C; Rosada, A; Avino, P

    2010-06-15

    This case study describes the characterization of a suspect rod of nuclear fuel seized in Italy: on request of the coroner, the characterization concerned the kind and the conditions of the rod, the amount and the specific characteristics of the species present in it, with particular attention to their possible use chemical and/or nuclear plants. The methodology used was based on radiochemical analyses (gammagraphic and gamma-spectrometry) whereas the comparison was performed by means of a fuel reference element working in the TRIGA nuclear reactor at Research Center of ENEA-Casaccia. The results show clearly how the exhibit was an element of nuclear fuel, how long it was irradiated, and the amount of (239)Pu produced and the (235)U consumed. Finally, even if the seized rod was briefly radiated at the "zero power" and traces of fission products and plutonium were found, it would be still usable as "fresh" fuel in a reactor type TRIGA if it had not been intercepted by Italian police authorities. PMID:20223608

  4. An investigation towards real time dose rate monitoring, and fuel rod detection in a First Generation Magnox Storage Pond (FGMSP).

    PubMed

    Jackson, Sarah F; Monk, Stephen D; Riaz, Zahid

    2014-12-01

    The First Generation Magnox Storage Pond (FGMSP) is located on the Sellafield Nuclear Site, housing legacy spent Magnox nuclear fuel. Some of which has since corroded, forming a layer of Corroded Magnox Sludge (CMS) creating one of the largest decommissioning challenges the UK has faced. In this work the composition, physical properties and potentially high hazard nature of CMS are discussed, as are the gamma emission spectra of spent Magnox fuel rods typical of the ilk stored. We assess the potential use of a RadLine gamma detector to dose rate map this area and provide fuel rod detection. RadLine consists of a small scintillator, fibre optic cable and photon counter. The probe has the unusual advantage of not being electrically active and therefore fully submersible underwater, with the option to deploy hundreds of metres in length. Our experimental method encompasses general purpose Monte Carlo radiation transport code, MCNP, where we describe the modelling of CMS and pond liquor in comprehensive detail, including their radiological spectrum, chemical composition data, and physical properties. This investigation concludes that the maximum energy deposited within the scintillator crystal due to ambient CMS corresponds to a dose rate of 5.65Gy h(-1), thus above this value positive detection of a fuel rod would be anticipated. It is additionally established that the detectable region is within a 20cm range. PMID:25244071

  5. Irradiation performance of long-rod duplex fuel-pellet bundle test - LDR test (AWBA Development Program)

    SciTech Connect

    Waldman, L.A.; Sphar, C.D.; Alff, T.H.

    1982-04-01

    The Long Duplex Rod (LDR) test is a five-rod bundle irradiation test of 0.30-inch diameter, 8-foot long, Zircaloy-clad rods containing duplex fuel pellets. These pellets include fissile material in an outer annulus surrounding fertile material in an inner cylindrical core. The rods are axially supported by top or bottom base plates and are laterally supported by an AM-350 stainless steel grid system. Design of the test, which includes duplex pellet annuli of three different compositions (UO/sub 2/, ZrO/sub 2/-UO/sub 2/-CaO, and ThO/sub 2/-UO/sub 2/), is described; and results of nondestructive examination after operation at peak linear power output of 9 to 18 Kw/ft to a peak depletion of 15 x 10/sup 20/ fissions/cm/sup 3/ of compartment (6.0 x 10/sup 4/ MWD/Tonne U + Th), or 30 x 10/sup 20/ fissions/cm/sup 3/ of fuel annulus volume (1.2 x 10/sup 5/ MWD/Tonne U + Th), are presented. It is concluded that performance capability of the duplex pellets is satisfactory for use in prebreeder reactor cores.

  6. Power Burst Facility: U(18)O2-CaO-ZrO2 Fuel Rods in Water

    SciTech Connect

    Jose Ignacio Marquez Damian; Alexis Weir; Valeria L. Putnam; John D. Bess

    2009-09-01

    The Power Burst Facility (PBF) reactor operated from 1972 to 1985 on the SPERT Area I of the Idaho National Laboratory, then known as Nuclear Reactor Test Station. PBF was designed to provide experimental data to aid in defining thresholds for and modes of failure under postulated accident conditions. PBF reactor startup testing began in 1972. This evaluation focuses on two operational loading tests, chronologically numbered 1 and 2, published in a startup-test report in 1974 [1]. Data for these tests was used by one of the authors to validate a MCNP model for criticality safety purposes [2]. Although specific references to original documents are kept in the text, all the reactor parameters and test specific data presented here was adapted from that report. The tests were performed with operational fuel loadings, a stainless steel in-pile tube (IPT) mockup, a neutron source, four pulse chambers, two fission chambers, and one ion chamber. The reactor's four transition rods (TRs) and control rods (CRs) were present but TR boron was completely withdrawn below the core and CR boron was partially withdrawn above the core. Test configurations differ primarily in the number of shim rods, and consequently the number of fuel rods included in the core. The critical condition was approached by incrementally and uniformly withdrawing CR boron from the core. Based on the analysis of the experimental data and numerical calculations, both experiments are considered acceptable as criticality safety benchmarks.

  7. Analysis of the dose rate produced by control rods discharged from a BWR into the irradiated fuel pool.

    PubMed

    Ródenas, J; Gallardo, S; Abarca, A; Juan, V

    2010-01-01

    BWR control rods become activated by neutron reactions into the reactor. Therefore, when they are withdrawn from the reactor, they must be stored into the storage pool for irradiated fuel at a certain depth under water. Dose rates on the pool surface and the area surrounding the pool should be lower than limits for workers. The MCNP code based on the Monte Carlo method has been applied to model this situation and to calculate dose rates at points of interest. PMID:19836252

  8. Mechanistic modeling of evaporating thin liquid film instability on a BWR fuel rod with parallel and cross vapor flow

    NASA Astrophysics Data System (ADS)

    Hu, Chih-Chieh

    This work has been aimed at developing a mechanistic, transient, 3-D numerical model to predict the behavior of an evaporating thin liquid film on a non-uniformly heated cylindrical rod with simultaneous parallel and cross flow of vapor. Interest in this problem has been motivated by the fact that the liquid film on a full-length boiling water reactor fuel rod may experience significant axial and azimuthal heat flux gradients and cross flow due to variations in the thermal-hydraulic conditions in surrounding subchannels caused by proximity to inserted control blade tip and/or the top of part-length fuel rods. Such heat flux gradients coupled with localized cross flow may cause the liquid film on the fuel rod surface to rupture, thereby forming a dry hot spot. These localized dryout phenomena can not be accurately predicted by traditional subchannel analysis methods in conjunction with empirical dryout correlations. To this end, a numerical model based on the Level Contour Reconstruction Method was developed. The Standard k-ε turbulence model is included. A cylindrical coordinate system has been used to enhance the resolution of the Level Contour Reconstruction Model. Satisfactory agreement has been achieved between the model predictions and experimental data. A model of this type is necessary to supplement current state-of-the-art BWR core thermal-hydraulic design methods based on subchannel analysis techniques coupled with empirical dry out correlations. In essence, such a model would provide the core designer with a "magnifying glass" by which the behavior of the liquid film at specific locations within the core (specific axial node on specific location within a specific bundle in the subchannel analysis model) can be closely examined. A tool of this type would allow the designer to examine the effectiveness of possible design changes and/or modified control strategies to prevent conditions leading to localized film instability and possible fuel failure.

  9. Data summary report for the destructive examination of Rods G7, G9, J8, I9, and H6 from Turkey Point Fuel Assembly B17

    SciTech Connect

    Davis, R B; Pasupathi, V

    1981-04-01

    Destructive examination results of five spent fuel rods from a Turkey Point Unit 3 pressurized water reactor are reported. Examinations included fission gas analysis, cladding hydrogen content analysis, fuel burnup analysis, metallographic examination, autoradiography and shielded electron microprobe analysis. All rods were found to be of sound integrity with an average burnup of 27 GWd/MTU and a 0.3% fission gas release.

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

    SciTech Connect

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

    1997-12-01

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

  11. Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

    DOE PAGESBeta

    Wang, Hong; Wang, Jy-An John

    2016-07-20

    We studied behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending. Tests were performed under load or moment control at 5 Hz, and an empirical correlation was established between rod fatigue life and amplitude of the applied moment. Fatigue response of Zry-4 cladding was further characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment applied and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition all affect surrogate rod failure. Bonding/debonding of PPI/PCI and pellet fracturing contribute to surrogatemore » rod bending fatigue. Also, the effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective specimen gauge length is effective in sensor spacing correction. Finally, we developed the database and gained understanding in this study such that it will serve as input to analysis of SNF vibration integrity.« less

  12. Numerical design of outer diameter remote field eddy current probe for the inspection of nuclear fuel rod

    NASA Astrophysics Data System (ADS)

    Shin, Young-Kil; Sun, Yushi

    2001-04-01

    In this paper, an encircling outer diameter (OD) remote field eddy current (RFEC) probe is proposed to inspect the nuclear fuel rod. To force the electromagnetic energy from exciter coil to penetrate into the rod, shielding by laminations of iron is applied outside the exciter coil. The operating frequency and effects of shielding are studied by the finite element analysis and modeling results show the existence of RFEC effects. Based on these results, the location for an encircling OD sensor coil is decided. However, predicted signals do not clearly show defect indications when the sensor passes a defect and exhibit certain symptoms that the fields from the exciter directly affect the sensor signal. To prevent direct contact with exciter fields, the sensor coil is also shielded. This shielding of sensor coil dramatically improves signal characteristics. The resulting signals have very similar characteristics to those of inner diameter RFEC signals and show almost equal sensitivity to inner diameter and outer diameter defects.

  13. High temperature post-irradiation performance of spent pressurized-water-reactor fuel rods under dry-storage conditions

    SciTech Connect

    Einziger, R.E.; Atkin, S.D.; Stellrecht, D.E.; Pasupathi, V.

    1981-06-01

    Post-irradiation studies on failure mechanisms of well characterized PWR rods were conducted for up to a year at 482, 510 and 571/sup 0/C in unlimited air and inert gas atmospheres. No cladding breaches occurred even though the tests operated many orders of magnitude longer in time than the lifetime predicted by Blackburn's analyses. The extended lifetime is due to significant creep strain of the Zircaloy cladding which decreases the internal rod pressures. The cladding creep also contributes to radial cracks, through the external oxide and internal FCCI layers, which propagated into and arrested in an oxygen stabilized ..cap alpha..-Zircaloy layer. There were no signs of either additional cladding hydriding, stress-corrosion cracking or fuel pellet degradation. Using the Larson-Miller formulization, a conservative maximum storage temperature of 400/sup 0/C is recommended to ensure a 1000-year cladding lifetime. This accounts for crack propagation and assumes annealing of the irradiation-hardened cladding.

  14. Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd/t

    SciTech Connect

    Caruso, S.; Murphy, M.; Jatuff, F.; Chawla, R.

    2006-07-01

    High-resolution gamma spectroscopy has been employed for the measurement of {sup 134}Cs/{sup 137}Cs, {sup 154}Eu/{sup 137}Cs and {sup 134}Cs/{sup 154}Eu gamma intensity ratios from spent fuel with the purpose of deriving pin-averaged single-ratio burnup indicators for high and ultra-high burnups. Two UO{sub 2} pressurised water reactor (PWR) fuel rod segments with record burnup levels >80 GWd/t have been experimentally characterised. Additionally, pin cell depletion calculations have been performed for each sample with the deterministic code CASMO-4, using both its JEF2.2- and its ENDF/B-IV-based libraries, for three different descriptions of the fuel rod irradiation histories, in order to test the sensitivity of the results to neutron cross sections and to the depletion model employed. Measured and calculated ratios have then been compared. It is shown that the {sup 134}Cs/{sup 137}Cs ratio, frequently used as burnup monitor, is considerably less accurate for values exceeding 50 GWd/t; discrepancies of up to {approx}25% are found between measured and calculated values. The ratios built with the {sup 154}Eu concentration show much larger discrepancies, essentially because this isotope is rather poorly predicted as revealed by the use of different basic cross section data. (authors)

  15. Analysis of Actual Operating Conditions of an Off-grid Solid Oxide Fuel Cell

    SciTech Connect

    Dennis Witmer; Thomas Johnson; Jack Schmid

    2008-12-31

    Fuel cells have been proposed as ideal replacements for other technologies in remote locations such as Rural Alaska. A number of suppliers have developed systems that might be applicable in these locations, but there are several requirements that must be met before they can be deployed: they must be able to operate on portable fuels, and be able to operate with little operator assistance for long periods of time. This project was intended to demonstrate the operation of a 5 kW fuel cell on propane at a remote site (defined as one without access to grid power, internet, or cell phone, but on the road system). A fuel cell was purchased by the National Park Service for installation in their newly constructed visitor center at Exit Glacier in the Kenai Fjords National Park. The DOE participation in this project as initially scoped was for independent verification of the operation of this demonstration. This project met with mixed success. The fuel cell has operated over 6 seasons at the facility with varying degrees of success, with one very good run of about 1049 hours late in the summer of 2006, but in general the operation has been below expectations. There have been numerous stack failures, the efficiency of electrical generation has been lower than expected, and the field support effort required has been far higher than expected. Based on the results to date, it appears that this technology has not developed to the point where demonstrations in off road sites are justified.

  16. Thermal expansion of UO2+x nuclear fuel rods from a model coupling heat transfer and oxygen diffusion

    SciTech Connect

    Mihaila, Bogden; Zubelewicz, Aleksander; Stan, Marius; Ramirez, Juan

    2008-01-01

    We study the thermal expansion of UO{sub 2+x} nuclear fuel rod in the context of a model coupling heat transfer and oxygen diffusion discussed previously by J.C. Ramirez, M. Stan and P. Cristea [J. Nucl. Mat. 359 (2006) 174]. We report results of simulations performed for steady-state and time-dependent regimes in one-dimensional configurations. A variety of initial- and boundary-value scenarios are considered. We use material properties obtained from previously published correlations or from analysis of previously published data. All simulations were performed using the commercial code COMSOL Multiphysics{sup TM} and are readily extendable to include multidimensional effects.

  17. FEMAXI-V benchmarking study on peak temperature and fission gas release prediction of PWR rod fuel

    SciTech Connect

    Suwardi; Dewayatna, W.; Briyatmoko, B.

    2012-06-06

    The present paper reports a study of FEMAXI-V code and related report on code benchmarking. Capabilities of the FEMAXI-V code to predict the thermal and fission gas release have been tested on MOX fuels in LWRs which has been done in SCK{center_dot}CEN and Belgonucleaire by using PRIMO MOX rod BD8 irradiation experiment after V Sobolev as reported O. J. Ott. Base irradiation in the BR3 reactor, the BD8 rod was transported to CEA-Saclay for irradiation in the OSIRIS reactor (ramp power excursion). The irradiation device used for the PRIMO ramps was the ISABELLE 1 loop, installed on a movable structure of the core periphery. The power variations were obtained by inwards/backwards movements of the loop in the core water. The preconditioning phase for rod BD8 occurred at a peak power level of 189 W/cm with a hold time of 27 hours. The subsequent power excursion rate amounted to 77 W/ (cm.min), reaching a terminal peak power level of 395 W/cm that lasted for 20 hours.

  18. 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

  19. Sustainable Power Generation in Continuous Flow Microbial Fuel Cell Treating Actual Wastewater: Influence of Biocatalyst Type on Electricity Production

    PubMed Central

    Ismail, Zainab Z.; Jaeel, Ali Jwied

    2013-01-01

    Microbial fuel cells (MFCs) have the potential to simultaneously treat wastewater for reuse and to generate electricity. This study mainly considers the performance of an upflow dual-chambered MFC continuously fueled with actual domestic wastewater and alternatively biocatalyzed with aerobic activated sludge and strain of Bacillus Subtilis. The behavior of MFCs during initial biofilm growth and characterization of anodic biofilm were studied. After 45 days of continuous operation, the biofilms on the anodic electrode were well developed. The performance of MFCs was mainly evaluated in terms of COD reductions and electrical power output. Results revealed that the COD removal efficiency was 84% and 90% and the stabilized power outputs were clearly observed achieving a maximum value of 120 and 270 mW/m2 obtained for MFCs inoculated with mixed cultures and Bacillus Subtilis strain, respectively. PMID:24453893

  20. Modeling of chemical interactions of fuel rod materials at high temperatures II. Investigation of downward relocation of molten materials

    NASA Astrophysics Data System (ADS)

    Veshchunov, M. S.; Palagin, A. V.

    1998-01-01

    In Part II of the modeling of chemical interactions of fuel rod materials at high temperatures, qualitative results on the nature of Zr-rich melt oxidation and interactions with fuel rods allow further interpretation of the post-test examinations of structures (debris) formed in the CORA tests under more complicated conditions, namely during downward relocation of the melt. In this situation, the molten mass extensively oxidizes and simultaneously dissolves UO 2 pellets and ZrO 2 scales of the cladding. The analysis of these simultaneous physico-chemical processes on the basis of the kinetic oxidation/dissolution model developed in Part I of the paper, allows a new interpretation and explanation of the CORA tests results concerning relocation dynamics of the major part of the melt (slow relocation of melt in the form of massive slug rather than quick relocations of droplets and rivulets), formation of local blockages (debris) in the interrod space and accumulation of the melt in the core region in the form of molten pool.

  1. Irradiation performance of HTGR fuel rods in HFIR experiments HRB-11 and -12

    SciTech Connect

    Homan, F.J.; Tiegs, T.N.; Kania, M.J.; Long, E.L. Jr.; Thoms, K.R.; Robbins, J.M.; Wagner, P.

    1980-06-01

    Capsules HRB-11 and -12 were irradiated in support of development of weak-acid-resin-derived recycle fuel for the high-enriched uranium (HEU) fuel cycle for the HTGR. Fissil fuel particles with initial oxygen-to-metal ratios between 1.0 and 1.7 performed acceptably to full burnup for HEU fuel. Particles with ratios below 1.0 showed excessive chemical interaction between rare earth fission products and the SiC layer.

  2. On the actual cathode mixed potential in direct methanol fuel cells

    NASA Astrophysics Data System (ADS)

    Zago, M.; Bisello, A.; Baricci, A.; Rabissi, C.; Brightman, E.; Hinds, G.; Casalegno, A.

    2016-09-01

    Methanol crossover is one of the most critical issues hindering commercialization of direct methanol fuel cells since it leads to waste of fuel and significantly affects cathode potential, forming a so-called mixed potential. Unfortunately, due to the sluggish anode kinetics, it is not possible to obtain a reliable estimation of cathode potential by simply measuring the cell voltage. In this work we address this limitation, quantifying the mixed potential by means of innovative open circuit voltage (OCV) tests with a methanol-hydrogen mixture fed to the anode. Over a wide range of operating conditions, the resulting cathode overpotential is between 250 and 430 mV and is strongly influenced by methanol crossover. We show using combined experimental and modelling analysis of cathode impedance that the methanol oxidation at the cathode mainly follows an electrochemical pathway. Finally, reference electrode measurements at both cathode inlet and outlet provide a local measurement of cathode potential, confirming the reliability of the innovative OCV tests and permitting the evaluation of cathode potential up to typical operating current. At 0.25 A cm-2 the operating cathode potential is around 0.85 V and the Ohmic drop through the catalyst layer is almost 50 mV, which is comparable to that in the membrane.

  3. Processing of the GALILEOTM fuel rod code model uncertainties within the AREVA LWR realistic thermal-mechanical analysis methodology

    NASA Astrophysics Data System (ADS)

    Mailhe, P.; Barbier, B.; Garnier, Ch.; Landskron, H.; Sedlacek, R.; Arimescu, I.; Smith, M.; Bellanger, Ph.

    2014-06-01

    The availability of reliable tools and associated methodology able to accurately predict the LWR fuel behavior in all conditions is of great importance for safe and economic fuel usage. For that purpose, AREVA has developed its new global fuel rod performance code GALILEOTM along with its associated realistic thermal-mechanical analysis methodology. This realistic methodology is based on a Monte Carlo type random sampling of all relevant input variables. After having outlined the AREVA realistic methodology, this paper will be focused on the GALILEOTM code benchmarking process on its extended experimental database and the GALILEOTM model uncertainties assessment. The propagation of these model uncertainties through the AREVA realistic methodology is also presented. This GALILEOTM model uncertainties processing is of the utmost importance for accurate fuel design margin evaluation as illustrated on some application examples. With the submittal of Topical Report for GALILEOTM to the U.S. NRC in 2013, GALILEOTM and its methodology are on the way to be industrially used in a wide range of irradiation conditions.

  4. Rodding Surgery

    MedlinePlus

    ... Rods can be made of stainless steel or titanium. Regular rods do not expand. They have many ... v regular), the rod materials (stainless steel v titanium) and the age for a first rodding surgery. ...

  5. Irradiation of SiC Clad Fuel Rods in the HFIR

    SciTech Connect

    Ott, Larry J; Bell, Gary L; Ellis, Ronald James; McDuffee, Joel Lee; Morris, Robert Noel

    2013-01-01

    During 2009 and- 2010, new test capability for the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was developed that allows testing of advanced nuclear fuels and cladding under prototypic light-water-reactor (LWR) operating conditions (i.e., cladding and fuel temperatures, fuel average linear heat generation rates, and cladding fluence). For the initial experiments for this test program, ORNL teamed with commercial fuel/cladding vendors who have developed an advanced composite-wound SiC cladding material for possible use in LWRs. The first experiment, containing SiC-clad UN fuel, was inserted in HFIR in June 2010, and the second experiment, containing SiC-clad UO2 fuel, was inserted in October 2010. Two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in November 2011 at an estimated fuel burnup of approximately 10 GWd/MTHM; and two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in February 2013 at an estimated fuel burnup of approximately 20 GWd/MTHM. These capsules are currently awaiting PIE. This paper will describe the experiment, as-run operating conditions for these capsules, and current PIE plans and status.

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

    SciTech Connect

    Chris Newman; Glen Hansen; Derek Gaston

    2009-07-01

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

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

    SciTech Connect

    Donald Olander

    2005-08-24

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

  8. Limits on the experimental simulation of nuclear fuel rod response. [PWR

    SciTech Connect

    Hagar, R.C.

    1980-01-01

    The steady-state and transient effects of intrinxic geometric and material property differences between typical nuclear fuel pins and electric fuel pin simulators (FPSs) are identified. The effectiveness of varying the transient power supplied to the FPS in reducing the differences between the transient responses of nuclear fuel pins and FPSs is investigated. This effectiveness is shown to be limited by the heat capacity of the FPS, the allowed range of the power program, and different FPS power requirements at different positions on a full-length FPS.

  9. Spot weld attachment of thermocouples to a fuel rod cladding interior surface

    SciTech Connect

    Page, R.E.; Bates, S.O.; Pilger, J.P.

    1984-08-01

    Research was conducted by Pacific Northwest Laboratory to weld 0.020-inch-diameter thermocouples to the interior surface of Zircaloy 4 light-water reactor fuel cladding. Inconel sheathed Type K thermocouples were attached to fuel cladding to register cladding temperatures during loss-of-coolant accident testing. This report describes the development of welding parameters and the effects of thermocouple attachment on the burst strength and integrity of the cladding at temperatures up to 1550/sup 0/F.

  10. Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

    SciTech Connect

    Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D.

    1997-08-01

    Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.

  11. Stochastic simulation of fission product activity in primary coolant due to fuel rod failures in typical PWRs under power transients

    NASA Astrophysics Data System (ADS)

    Javed Iqbal, M.; Mirza, Nasir M.; Mirza, Sikander M.

    2008-01-01

    During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the 131I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent 135I, 135Xe and 89Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.

  12. Evaluation of alternative treatments for spent fuel rod consolidation wastes and other miscellaneous commercial transuranic wastes

    SciTech Connect

    Ross, W.A.; Schneider, K.J.; Oma, K.H.; Smith, R.I.; Bunnell, L.R.

    1986-05-01

    Eight alternative treatments (and four subalternatives) are considered for both existing commercial transuranic wastes and future wastes from spent fuel consolidation. Waste treatment is assumed to occur at a hypothetical central treatment facility (a Monitored Retrieval Storage facility was used as a reference). Disposal in a geologic repository is also assumed. The cost, process characteristics, and waste form characteristics are evaluated for each waste treatment alternative. The evaluation indicates that selection of a high-volume-reduction alternative can save almost $1 billion in life-cycle costs for the management of transuranic and high-activity wastes from 70,000 MTU of spent fuel compared to the reference MRS process. The supercompaction, arc pyrolysis and melting, and maximum volume reduction alternatives are recommended for further consideration; the latter two are recommended for further testing and demonstration.

  13. FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code

    SciTech Connect

    Valtonen, Keijo; Hamalainen, Anitta; Cunningham, Mitchel E. )

    2002-05-01

    The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.

  14. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods; Revision 1

    SciTech Connect

    Johnson, G.L.

    1991-11-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degrees}C and whether the cladding of the stored spent fuel ever exceeds 350{degrees}C. Limiting the borehole to temperatures of 97{degrees}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degrees}C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 {times} 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degrees}C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350{degrees}C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft {times} 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40{degrees}C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation.

  15. Multidimensional effects in the thermal response of fuel rod simulators. [PWR

    SciTech Connect

    Dabbs, R.D.; Ott, L.J.

    1980-01-01

    One of the primary objectives of the Oak Ridge National Laboratory Pressurized-Water Reactor Blowdown Heat Transfer Separate-Effects Program is the determination of the transient surface temperature and surface heat flux of fuel pin simulators (FPSs) from internal thermocouple signals obtained during a loss-of-coolant experiment (LOCE) in the Thermal-Hydraulics Test Facility. This analysis requires the solution of the classical inverse heat conduction problem. The assumptions that allow the governing differential equation to be reduced to one dimension can introduce significant errors in the computed surface heat flux and surface temperature. The degree to which these computed variables are perturbed is addressed and quantified.

  16. Forming Limit Diagrams of Zircaloy-4 and Zirlo Sheets for Stamping Process of Spacer Grids of Nuclear Fuel Rod

    NASA Astrophysics Data System (ADS)

    Seo, Yunmi; Hyun, Hong Chul; Lee, Hyungyil; Kim, Naksoo

    2011-08-01

    We investigated the theoretical forming limit models for Zircaloy-4 and Zirlo used for spacer grid of nuclear fuel rods. Tensile tests were performed to obtain stress-strain curves and anisotropic coefficients, such as r-values. The experimental forming limit diagrams (FLD) for two materials were obtained by dome stretching tests following the specification of NUMISHEET 96. Theoretical FLD depends on forming limit model and yield criterion. To obtain the right hand side of FLD, we applied the forming limit models (Swift's diffuse necking, Marciniak-Kuczynski damage defect, Storen-Rice's vertex theory) to Zircaloy-4 and Zirlo sheets. Hill's local necking theory was adopted for the left side of FLD. To consider the anisotropy of sheets, the yield criteria of Hill (1948) and Hosford (1979) were applied. Comparing the predicted curves with the experimental data, we found that the FLD for Zircaloy-4 can be described by the Swift model with the Hill 48 yield criterion, while the FLD for Zirlo can be explained by the Storen-Rice model and the Hosford yield criterion (a = 8).

  17. Anisotropic Azimuthal Power and Temperature distribution on FuelRod. Impact on Hydride Distribution

    SciTech Connect

    Motta, Arthur; Ivanov, Kostadin; Arramova, Maria; Hales, Jason

    2015-04-29

    The degradation of the zirconium cladding may limit nuclear fuel performance. In the high temperature environment of a reactor, the zirconium in the cladding corrodes, releasing hydrogen in the process. Some of this hydrogen is absorbed by the cladding in a highly inhomogeneous manner. The distribution of the absorbed hydrogen is extremely sensitive to temperature and stress concentration gradients. The absorbed hydrogen tends to concentrate near lower temperatures. This hydrogen absorption and hydride formation can cause cladding failure. This project set out to improve the hydrogen distribution prediction capabilities of the BISON fuel performance code. The project was split into two primary sections, first was the use of a high fidelity multi-physics coupling to accurately predict temperature gradients as a function of r, θ , and z, and the second was to use experimental data to create an analytical hydrogen precipitation model. The Penn State version of thermal hydraulics code COBRA-TF (CTF) was successfully coupled to the DeCART neutronics code. This coupled system was verified by testing and validated by comparison to FRAPCON data. The hydrogen diffusion and precipitation experiments successfully calculated the heat of transport and precipitation rate constant values to be used within the hydrogen model in BISON. These values can only be determined experimentally. These values were successfully implemented in precipitation, diffusion and dissolution kernels that were implemented in the BISON code. The coupled output was fed into BISON models and the hydrogen and hydride distributions behaved as expected. Simulations were conducted in the radial, axial and azimuthal directions to showcase the full capabilities of the hydrogen model.

  18. Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report

    SciTech Connect

    Siefken, L.J.

    1999-02-01

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ``Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

  19. 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

  20. 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

  1. Status of rod consolidation, 1988

    SciTech Connect

    Bailey, W.J.

    1989-01-01

    It is estimated that the spent fuel storage pools at some domestic light-water reactors will run out of space before 2003, the year that the US Department of Energy currently predicts it will have a repository available. Of the methods being studied to alleviate the problem, rod consolidation is one of the leading candidates for achieving more efficient use of existing space in spent fuel storage pools. Rod consolidation involves mechanically removing all the fuel rods from the fuel assembly hardware (i.e., the structural components) and placing the fuel rods in a close-packed array in a canister without space grids. A typical goal of rod consolidation systems is to insert the fuel rods from two fuel assemblies into a canister that has the same exterior dimensions as one standard fuel assembly (i.e., to achieve a consolidation or compaction ratio of 2:1) and to compact the nonfuel-bearing structural components from those two fuel assemblies by a factor of 10 to 20. This report provides an overview of the current status of rod consolidation in the United States and a small amount of information on related activities in other countries. 85 refs., 36 figs., 5 tabs.

  2. Processing, packaging, and storage of non-fuel-bearing components from the rod consolidation demonstration at Prairie Island Nuclear Generating Station

    SciTech Connect

    McCarten, L.; Kapitz, J.; Kaczmarsky, M.; Rec, J.

    1988-01-01

    Many nuclear power plants are running out of space in their spent-fuel pools, and by the early 1990s, existing spent-fuel storage capacity must be supplemented at over 20 commercial nuclear plants. Rod consolidation and dry storage, either individually or in combination, are the only viable alternatives to meet the spent-fuel storage requirements until a government storage facility or repository is established. The Prairie Island Nuclear Generating Station operated by Northern States Power Company (NSP) is in this predicament. To meet Prairie Island's storage needs, NSP is evaluating the feasibility of full-scale implementation of spent-fuel consolidation. The technical and economic success of fuel consolidation requires successful and economical processing, storage and disposal of the scrap non-fuel-bearing components (NFBC). In the fall of 1987, NSP initiated a consolidation demonstration program at Prairie Island, during which 29 equipment spent-fuel assemblies were successfully consolidated by Westinghouse. The paper discusses program scope, NFBC characterization and classification, NFBC processing and NFBC segregation and packaging.

  3. Rod consolidation at the West Valley Demonstration Project

    SciTech Connect

    Bailey, W.J.

    1986-12-01

    A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab.

  4. Reactor control rod timing system

    DOEpatents

    Wu, Peter T. K.

    1982-01-01

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

  5. APPARATUS FOR SHEATHING RODS

    DOEpatents

    Ford, W.K.; Wyatt, M.; Plail, S.

    1961-08-01

    An arrangement is described for sealing a solid body of nuclear fuel, such as a uranium metal rod, into a closelyfitting thin metallic sheath with an internal atmosphere of inert gas. The sheathing process consists of subjecting the sheath, loaded with the nuclear fuel body, to the sequential operations of evacuation, gas-filling, drawing (to entrap inert gas and secure close contact between sheath and body), and sealing. (AEC)

  6. A particle assembly/constrained expansion (PACE) model for the formation and structure of porous metal oxide deposits on nuclear fuel rods in pressurized light water reactors

    NASA Astrophysics Data System (ADS)

    Brenner, Donald W.; Lu, Shijing; O'Brien, Christopher J.; Bucholz, Eric W.; Rak, Zsolt

    2015-02-01

    A new model is proposed for the structure and properties of porous metal oxide scales (aka Chalk River Unidentified Deposits (CRUD)) observed on the nuclear fuel rod cladding in Pressurized Water Reactors (PWR). The model is based on the thermodynamically-driven expansion of agglomerated octahedral nickel ferrite particles in response to pH and temperature changes in the CRUD. The model predicts that porous nickel ferrite with internal {1 1 1} surfaces is a thermodynamically stable structure under PWR conditions even when the free energy of formation of bulk nickel ferrite is positive. This explains the pervasive presence of nickel ferrite in CRUD, observed CRUD microstructures, why CRUD maintains its porosity, and variations in porosity within the CRUD observed experimentally. This model is a stark departure from decades of conventional wisdom and detailed theoretical analysis of CRUD chemistry, and defines new research directions for model validation, and for understanding and ultimately controlling CRUD formation.

  7. Performance of high-velocity oxy-fuel-sprayed chromium carbide-nickel chromium coating in an actual boiler environment of a thermal power plant

    SciTech Connect

    Sidhu, T.S.; Prakash, S.; Agrawal, R.D.

    2007-09-15

    The present study aims to evaluate the performance of a high-velocity oxy-fuel (HVOF)-sprayed Cr{sub 3}C{sub 2}-NiCr (chromium carbide-nickel chromium) coating on a nickel-based super-alloy in an actual industrial environment of a coal-fired boiler, with the objective to protect the boiler super-heater and reheater tubes from hot corrosion. The tests were performed in the platen super heater zone of a coal-fired boiler for 1,000 h at 900 degrees C under cyclic conditions. The Cr{sub 3}C{sub 2}-NiCr coating imparted the necessary protection to the nickel-based super alloy in the given environment. The dense and flat splat structure of the coating, and the formation of oxides of chromium and nickel and their spinels, might have protected the substrate super alloy from the inward permeation of corrosive species.

  8. Partially-reflected water-moderated square-piteched U(6.90)O2 fuel rod lattices with 0.67 fuel to water volume ratio (0.800 CM Pitch)

    SciTech Connect

    Harms, Gary A.

    2015-09-01

    The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5% 235U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90%)O2 fuel rods.

  9. Development and verification of NRC`s single-rod fuel performance codes FRAPCON-3 AND FRAPTRAN

    SciTech Connect

    Beyer, C.E.; Cunningham, M.E.; Lanning, D.D.

    1998-03-01

    The FRAPCON and FRAP-T code series, developed in the 1970s and early 1980s, are used by the US Nuclear Regulatory Commission (NRC) to predict fuel performance during steady-state and transient power conditions, respectively. Both code series are now being updated by Pacific Northwest National Laboratory to improve their predictive capabilities at high burnup levels. The newest versions of the codes are called FRAPCON-3 and FRAPTRAN. The updates to fuel property and behavior models are focusing on providing best estimate predictions under steady-state and fast transient power conditions up to extended fuel burnups (> 55 GWd/MTU). Both codes will be assessed against a data base independent of the data base used for code benchmarking and an estimate of code predictive uncertainties will be made based on comparisons to the benchmark and independent data bases.

  10. Monopolar fuel cell stack coupled together without use of top or bottom cover plates or tie rods

    NASA Technical Reports Server (NTRS)

    Narayanan, Sekharipuram R. (Inventor); Valdez, Thomas I. (Inventor)

    2009-01-01

    A monopolar fuel cell stack comprises a plurality of sealed unit cells coupled together. Each unit cell comprises two outer cathodes adjacent to corresponding membrane electrode assemblies and a center anode plate. An inlet and outlet manifold are coupled to the anode plate and communicate with a channel therein. Fuel flows from the inlet manifold through the channel in contact with the anode plate and flows out through the outlet manifold. The inlet and outlet manifolds are arranged to couple to the inlet and outlet manifolds respectively of an adjacent one of the plurality of unit cells to permit fuel flow in common into all of the inlet manifolds of the plurality of the unit cells when coupled together in a stack and out of all of the outlet manifolds of the plurality of unit cells when coupled together in a stack.

  11. Measurement of void volume of a fuel rod and the exchange of occluded gases from mixed carbide fuel with filling gas helium

    NASA Astrophysics Data System (ADS)

    Rao, G. A. Rama; Kukarni, S. G.; Venugoopal, V.; Manchanda, V. K.; Goswami, G. L.

    1995-02-01

    The presence of gaseous impurities in the filling gas of a fuel pin is detriental to the thermal performance of a nuclear reactor fuel. The composition of the filling gas does not remain constant throughout the life of the fuel pin. The gas exchange phenomena that occur between the cover gas and impurity gases affect the fuel performance more severely in (U, Pu)O 2 fuel pin due to its inherently poor thermal conductivity than in advanced fuels such as mixed carbides and nitrides. In the present study the exchange phenomenon of the occluded gases present in our Fast Breeder Test Reactor (FBTR) fuel pellets [(U 0.30, Pu 0.70)C with 6500 ppm o] with the cover gas helium was observed as a function of time and temperature. Quantitative analysis of the released gases namely H 2, O 2 + Ar, N 2, CH 4 and CO was carried out at subambient pressure by gas chromatography. The void volume of the fuel element is determined experimentally by gas equilibration with known volume.

  12. Morphological Analysis of Zirconium Nuclear Fuel Retaining Rods Braided with SiC: Quality Assurance and Defect Identification

    SciTech Connect

    Michael V Glazoff; Robert Hiromoto; Akira Tokuhiro

    2014-08-01

    In the after-Fukushima world, the stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Among the methods explored currently to improve zircaloys’ thermal stability in off-normal conditions, using a protective coat of the SiC filaments is considered because silicon carbide is well known for its remarkable chemical inertness at high temperatures. A typical SiC fiber contains ~50,000 individual filaments of 5 – 10 µm in diameter. In this paper, an effort was made to develop and apply mathematical morphology to the process of automatic defect identification in Zircaloy-4 rods braided with the protective layer of the silicon carbide filament. However, the issues of the braiding quality have to be addressed to ensure its full protective potential. We present the original mathematical morphology algorithms that allow solving this problem of quality assurance successfully. In nuclear industry, such algorithms are used for the first time, and could be easily generalized to the case of automated continuous monitoring for defect identification in the future.

  13. Morphological analysis of zirconium nuclear fuel retaining rods braided with SiC: Quality assurance and defect identification

    NASA Astrophysics Data System (ADS)

    Glazoff, Michael V.; Hiromoto, Robert; Tokuhiro, Akira

    2014-08-01

    In the after-Fukushima world, the stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Among the methods explored currently to improve zircaloys’ thermal stability in off-normal conditions, using a protective coat of the SiC filaments is considered because silicon carbide is well known for its remarkable chemical inertness at high temperatures. A typical SiC fiber contains ∼50,000 individual filaments of 5-10 μm in diameter. In this paper, an effort was made to develop and apply mathematical morphology to the process of automatic defect identification in Zircaloy-4 rods braided with the protective layer of the silicon carbide filament. However, the issues of the braiding quality have to be addressed to ensure its full protective potential. We present the original mathematical morphology algorithms that allow solving this problem of quality assurance successfully. In nuclear industry, such algorithms are used for the first time, and could be easily generalized to the case of automated continuous monitoring for defect identification in the future.

  14. Fuel bundle design for enhanced usage of plutonium fuel

    DOEpatents

    Reese, Anthony P.; Stachowski, Russell E.

    1995-01-01

    A nuclear fuel bundle includes a square array of fuel rods each having a concentration of enriched uranium and plutonium. Each rod of an interior array of the rods also has a concentration of gadolinium. The interior array of rods is surrounded by an exterior array of rods void of gadolinium. By this design, usage of plutonium in the nuclear reactor is enhanced.

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

    SciTech Connect

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

    1981-02-01

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

  16. CONTROL ROD

    DOEpatents

    Zinn, W.H.; Ross, H.V.

    1958-11-18

    A control rod is described for a nuclear reactor. In certaln reactor designs it becomes desirable to use a control rod having great width but relatively llttle thickness. This patent is addressed to such a need. The neutron absorbing material is inserted in a triangular tube, leaving volds between the circular insert and the corners of the triangular tube. The material is positioned within the tube by the use of dummy spacers to achleve the desired absorption pattern, then the ends of the tubes are sealed with suitable plugs. The tubes may be welded or soldered together to form two flat surfaces of any desired width, and covered with sheetmetal to protect the tubes from damage. This design provides a control member that will not distort under the action of outside forces or be ruptured by gases generated within the jacketed control member.

  17. Literature search for the non-aqueous separation of zinc from fuel rod cladding. [After dissolution in liquid metal

    SciTech Connect

    Sandvig, R. L.; Dyer, S. J.; Lambert, G. A.; Baldwin, C. E.

    1980-06-21

    This report reviews the literature of processes for the nonaqueous separation of zinc from dissolved fuel assembly cladding. The processes considered were distillation, pyrochemical processing, and electrorefining. The last two techniques were only qualitatively surveyed while the first, distillation, was surveyed in detail. A survey of available literature from 1908 through 1978 on the distillation of zinc was performed. The literature search indicated that a zinc recovery rate in excess of 95% is possible; however, technical problems exist because of the high temperatures required and the corrosive nature of liquid zinc. The report includes a bibliography of the surveyed literature and a computer simulation of vapor pressures in binary systems. 129 references.

  18. Characterization and simulation of soft gamma-ray mirrors for their use with spent fuel rods at reprocessing facilities.

    PubMed

    Ruz, J; Descalle, M A; Alameda, J B; Brejnholt, N F; Chichester, D L; Decker, T A; Fernandez-Perea, M; Hill, R M; Kisner, R A; Melin, A M; Patton, B W; Soufli, R; Trellue, H; Watson, S M; Ziock, K P; Pivovaroff, M J

    2016-06-01

    The use of a grazing incidence optic to selectively reflect K-shell fluorescence emission and isotope-specific lines from special nuclear materials is a highly desirable nondestructive analysis method for use in reprocessing fuel environments. Preliminary measurements have been performed, and a simulation suite has been developed to give insight into the design of the x ray optics system as a function of the source emission, multilayer coating characteristics, and general experimental configurations. The experimental results are compared to the predictions from our simulation toolkit to illustrate the ray-tracing capability and explore the effect of modified optics in future measurement campaigns. PMID:27411177

  19. Characteristics of fuel crud and its impact on storage, handling, and shipment of spent fuel. [Fuel crud

    SciTech Connect

    Hazelton, R.F.

    1987-09-01

    Corrosion products, called ''crud,'' form on out-of-reactor surfaces of nuclear reactor systems and are transported by reactor coolant to the core, where they deposit on external fuel-rod cladding surfaces and are activated by nuclear reactions. After discharge of spent fuel from a reactor, spallation of radioactive crud from the fuel rods could impact wet or dry storage operations, handling (including rod consolidation), and shipping. It is the purpose of this report to review earlier (1970s) and more recent (1980s) literature relating to crud, its characteristics, and any impact it has had on actual operations. Crud characteristics vary from reactor type to reactor type, reactor to reactor, fuel assembly to fuel assembly in a reactor, circumferentially and axially in an assembly, and from cycle to cycle for a specific facility. To characterize crud of pressurized-water (PWRs) and boiling-water reactors (BWRs), published information was reviewed on appearance, chemical composition, areal density and thickness, structure, adhesive strength, particle size, and radioactivity. Information was also collected on experience with crud during spent fuel wet storage, rod consolidation, transportation, and dry storage. From experience with wet storage, rod consolidation, transportation, and dry storage, it appears crud spallation can be managed effectively, posing no significant radiological problems. 44 refs., 11 figs.

  20. Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Final Design Report

    SciTech Connect

    Siefken, L.J.

    1999-05-01

    Final designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. A description is given of the implementation of the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5/MOD3.3 code.

  1. Calculation of Hydrogen and Oxygen Uptake in Fuel Rod Cladding During Severe Accidents Using the Integral Diffusion Method - Final Design Report

    SciTech Connect

    Siefken, Larry James

    1999-06-01

    Final designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. A description is given of the implementation of the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5/MOD3.3 code.

  2. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    SciTech Connect

    Webb, B.J.

    1988-01-01

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab.

  3. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods (1.506-cm Pitch)

    SciTech Connect

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed from 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.a 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 performed to determine critical reflector arrangements, relative 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” (see Reference 1). The experiment studied in this evaluation was the second of the series and had the fuel rods in a 1.506-cm-triangular pitch. One critical configuration was found (see Reference 3). Once the critical configuration had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U,bc and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configuration are described in Sections 1.3, 1.4, and 1.7, respectively.

  4. The Waste Package Project. Final report, July 1, 1995--February 27, 1996: Volume 1, The structural performance of the shell and fuel rods of a high level nuclear waste container

    SciTech Connect

    Ladkany, S.G.; Rajagopalan, R.

    1996-06-01

    This dissertation proposal covers research work that started in the spring of 1992. The aim of the research has been to study the structural performance and stability of proposed nuclear waste containers and the enclosed fuel rods to be used in the long term storage of High Level Nuclear Waste (HLNW). This research is in two phases, computational and experimental. The computational phase deals with the linear and nonlinear Finite Element Analysis of the different containers due to various loading conditions during normal handling conditions and due to the effect of long term corrosion while the canister is stored in the drift of a backfilled geological repository. The elastoplastic stability of the nuclear fuel rods were studied under body forces resulting from acceleration vectors at varying angles, resulting from a sudden drop of the canister at an angle onto a hard surface.

  5. Feasibility demonstration of using wire electrical-discharge machining, abrasive flow honing, and laser spot welding to manufacture high-precision triangular-pitch Zircaloy-4 fuel-rod-support grids

    SciTech Connect

    Horwood, W.A.

    1982-05-01

    Results are reported supporting the feasibility of manufacturing high precision machined triangular pitch Zircaloy-4 fuel rod support grids for application in water cooled nuclear power reactors. The manufacturing processes investigated included wire electrical discharge machining of the fuel rod and guide tube cells in Zircaloy plate stock to provide the grid body, multistep pickling of the machined grid to provide smooth and corrosion resistant surfaces, and laser welding of thin Zircaloy cover plates to both sides of the grid body to capture separate AM-350 stainless steel insert springs in the grid body. Results indicated that dimensional accuracy better than +- 0.001 and +- 0.002 inch could be obtained on cell shape and position respectively after wire EDM and surface pickling. Results on strength, corrosion resistance, and internal quality of laser spot welds are provided.

  6. Reactor control rod timing system. [LMFBR

    DOEpatents

    Wu, P.T.K.

    1980-03-18

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

  7. 13. SOUTHEAST TO SUCKER ROD WORK BENCH AND WOODEN SUCKER ...

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

    13. SOUTHEAST TO SUCKER ROD WORK BENCH AND WOODEN SUCKER ROD STORAGE RACKS ALONG EAST WALL OF FACTORY INTERIOR. AT THIS BENCH WORKERS RIVETED THREADED WROUGHT IRON CONNECTORS TO THE ENDS OF 20' LONG WOODEN SUCKER RODS (THE RODS WHICH EXTEND DOWNWARD IN THE WELL FROM THE GROUND SURFACE TO PISTON DISPLACEMENT PUMPS WHICH ACTUALLY ELEVATE WATER TO THE SURFACE). ROZNOR HEATER AT THE FAR RIGHT WAS ADDED CIRCA 1960. - Kregel Windmill Company Factory, 1416 Central Avenue, Nebraska City, Otoe County, NE

  8. Modeling of chemical interactions of fuel rod materials at high temperatures I. Simultaneous dissolution of UO 2 and ZrO 2 by molten Zr in an oxidizing atmosphere

    NASA Astrophysics Data System (ADS)

    Veshchunov, M. S.; Berdyshev, A. V.

    1998-01-01

    Investigations and modeling of high-temperature processes associated with the oxidation of UZrO molten mixtures under various conditions of severe accidents (intact heated fuel rods or relocating melt) are presented on the basis of a thorough analysis of metallographic post-test examination data obtained in the fuel bundle CORA experiments. In Part I a model of simultaneous dissolution of solid UO 2 and ZrO 2 phases by molten Zr in an oxidizing atmosphere (steam) is presented. This modeling is performed on the basis of the generalization of a previously developed model of UO 2 dissolution by molten Zr taking into account the chemical dissolution by convectively stirring melt of a ZrO 2 layer interacting with steam. The model describes the complicated kinetics of these interactions accompanied by precipitation of a ceramic (U, Zr)O 2- x phase in the bulk of the liquid. This process finally leads to complete conversion of the melt into the (U, Zr)O 2- x layer located between UO 2 fuel and ZrO 2 shell. Such a three-layer structure was regularly observed in the CORA post-test metallographic examinations of fuel rods heated above the melting point of Zr cladding ( T ≥ 1950°C).

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

    SciTech Connect

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

    1980-01-01

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

  10. Nuclear thermionic converter. [tungsten-thorium oxide rods

    NASA Technical Reports Server (NTRS)

    Phillips, W. M.; Mondt, J. F. (Inventor)

    1977-01-01

    Efficient nuclear reactor thermionic converter units are described which can be constructed at low cost and assembled in a reactor which requires a minimum of fuel. Each converter unit utilizes an emitter rod with a fluted exterior, several fuel passages located in the bulges that are formed in the rod between the flutes, and a collector receiving passage formed through the center of the rod. An array of rods is closely packed in an interfitting arrangement, with the bulges of the rods received in the recesses formed between the bulges of other rods, thereby closely packing the nuclear fuel. The rods are constructed of a mixture of tungsten and thorium oxide to provide high power output, high efficiency, high strength, and good machinability.

  11. Rod-Coil Block Polyimide Copolymers

    NASA Technical Reports Server (NTRS)

    Meador, Mary Ann B. (Inventor); Kinder, James D. (Inventor)

    2005-01-01

    This invention is a series of rod-coil block polyimide copolymers that are easy to fabricate into mechanically resilient films with acceptable ionic or protonic conductivity at a variety of temperatures. The copolymers consist of short-rigid polyimide rod segments alternating with polyether coil segments. The rods and coil segments can be linear, branched or mixtures of linear and branched segments. The highly incompatible rods and coil segments phase separate, providing nanoscale channels for ion conduction. The polyimide segments provide dimensional and mechanical stability and can be functionalized in a number of ways to provide specialized functions for a given application. These rod-coil black polyimide copolymers are particularly useful in the preparation of ion conductive membranes for use in the manufacture of fuel cells and lithium based polymer batteries.

  12. Variable flow control for a nuclear reactor control rod

    DOEpatents

    Carleton, Richard D.; Bhattacharyya, Ajay

    1978-01-01

    A variable flow control for a control rod assembly of a nuclear reactor that depends on turbulent friction though an annulus. The annulus is formed by a piston attached to the control rod drive shaft and a housing or sleeve fitted to the enclosure housing the control rod. As the nuclear fuel is burned up and the need exists for increased reactivity, the control rods are withdrawn, which increases the length of the annulus and decreases the rate of coolant flow through the control rod assembly.

  13. Fuel bundle design for enhanced usage of plutonium fuel

    DOEpatents

    Reese, A.P.; Stachowski, R.E.

    1995-08-08

    A nuclear fuel bundle includes a square array of fuel rods each having a concentration of enriched uranium and plutonium. Each rod of an interior array of the rods also has a concentration of gadolinium. The interior array of rods is surrounded by an exterior array of rods void of gadolinium. By this design, usage of plutonium in the nuclear reactor is enhanced. 10 figs.

  14. Control rod drive

    SciTech Connect

    Hawke, Basil C.

    1986-01-01

    A control rod drive uses gravitational forces to insert one or more control rods upwardly into a reactor core from beneath the reactor core under emergency conditions. The preferred control rod drive includes a vertically movable weight and a mechanism operatively associating the weight with the control rod so that downward movement of the weight is translated into upward movement of the control rod. The preferred control rod drive further includes an electric motor for driving the control rods under normal conditions, an electrically actuated clutch which automatically disengages the motor during a power failure and a decelerator for bringing the control rod to a controlled stop when it is inserted under emergency conditions into a reactor core.

  15. Piston rod seal

    DOEpatents

    Lindskoug, Stefan

    1984-01-01

    In a piston rod seal of the type comprising a gland through which the piston rod is passed the piston is provided with a sleeve surrounding the piston rod and extending axially so as to axially partly overlap the gland when the piston is in its bottom dead center position.

  16. Rebirth of a control rod at the Phenix power plant

    SciTech Connect

    De Carvalho, Corinne; Vignau, Bernard; Masson, Marc

    2007-07-01

    This paper outlines the operations involved in cleaning the control rod for the complementary shutdown system in the Phenix Power Plant, the French sodium-cooled fast reactor. The Phenix reactor is controlled by six control rods and a complementary shutdown system. The latter comprises a control rod and a mechanism maintaining the rod in position by means of an electromagnet. The electromagnet is continuously supplied with power and holds the rod control assembly in position by magnetisation on a plane circular surface made from pure iron. The bearing capacity of the mechanism on the rod was initially 80 daN with a rod weight of 26.3 daN. This deteriorated progressively over time. The bearing surface of the rod and the electromagnet became contaminated with a deposit of sodium oxides and metallic particles, thus creating an air gap. This reached a figure of 36 daN in 2005 and was deemed not to be sufficient to prevent the rod from dropping at the wrong time during reactor operation. The Power Plant thus decided to replace the rod mechanism in the reactor in an initial phase, followed by the control rod itself. As the Phenix Power Plant had no spare control rods left, they initiated a 'salvage' plan, over two stages, for the rod removed from the reactor and placed in the fuel storage drum: - Inspection of the bearing surface of the rod by means of a borescope to check whether the rod could be salvaged, - A cleaning operation on the bearing face and checks on the bearing capacity of the rod. The operation is subject to very stringent requirements: the rod must not be taken out of the sodium to ensure that it can be reused in the reactor. The operation must thus take place in the fuel storage drum where there are no facilities for such an operation and where operating conditions are very hostile: high temperatures (the sodium in the fuel storage drum is at a temperature of 150 deg. C, high dose rate (3 mGy/h on the bearing surface) and the bearing surface is submerged

  17. CONTROL ROD DRIVE

    DOEpatents

    Chapellier, R.A.; Rogers, I.

    1961-06-27

    Accurate and controlled drive for the control rod is from an electric motor. A hydraulic arrangement is provided to balance a piston against which a control rod is urged by the application of fluid pressure. The electric motor drive of the control rod for normal operation is made through the aforementioned piston. In the event scramming is required, the fluid pressure urging the control rod against the piston is relieved and an opposite fluid pressure is applied. The lack of mechanical connection between the electric motor and control rod facilitates the scramming operation.

  18. CRUCIFORM CONTROL ROD JOINT

    DOEpatents

    Thorp, A.G. II

    1962-08-01

    An invention is described which relates to nuclear reactor control rod components and more particularly to a joint between cruciform control rod members and cruciform control rod follower members. In one embodiment this invention provides interfitting crossed arms at adjacent ends of a control rod and its follower in abutting relation. This holds the members against relative opposite longitudinal movement while a compression member keys the arms against relative opposite rotation around a common axis. Means are also provided for centering the control rod and its follower on a common axis and for selectively releasing the control rod from its follower for the insertion of a replacement of the control rod and reuse of the follower. (AEC)

  19. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    SciTech Connect

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  20. Using Finite Model Analysis and Out of Hot Cell Surrogate Rod Testing to Analyze High Burnup Used Nuclear Fuel Mechanical Properties

    SciTech Connect

    Wang, Jy-An John; Jiang, Hao; Wang, Hong

    2014-07-01

    Based on a series of FEA simulations, the discussions and the conclusions concerning the impact of the interface bonding efficiency to SNF vibration integrity are provided in this report; this includes the moment carrying capacity distribution between pellets and clad, and the impact of cohesion bonding on the flexural rigidity of the surrogate rod system. As progressive de-bonding occurs at the pellet-pellet interfaces and at the pellet-clad interface, the load ratio of the bending moment carrying capacity gradually shifts from the pellets to the clad; the clad starts to carry a significant portion of the bending moment resistance until reaching the full de-bonding state at the pellet-pellet interface regions. This results in localized plastic deformation of the clad at the pellet-pellet-clad interface region; the associated plastic deformations of SS clad leads to a significant degradation in the stiffness of the surrogate rod. For instance, the flexural rigidity was reduced by 39% from the perfect bond state to the de-bonded state at the pellet-pellet interfaces.

  1. Regulatory perspective on incomplete control rod insertions

    SciTech Connect

    Chatterton, M.

    1997-01-01

    The incomplete control rod insertions experienced at South Texas Unit 1 and Wolf Creek are of safety concern to the NRC staff because they represent potential precursors to loss of shutdown margin. Even before it was determined if these events were caused by the control rods or by the fuel there was an apparent correlation of the problem with high burnup fuel. It was determined that there was also a correlation between high burnup and high drag forces as well as with rod drop time histories and lack of rod recoil. The NRC staff initial actions were aimed at getting a perspective on the magnitude of the problem as far as the number of plants and the amount of fuel that could be involved, as well as the safety significance in terms of shutdown margin. As tests have been performed and data has been analyzed the focus has shifted more toward understanding the problem and the ways to eliminate it. At this time the staff`s understanding of the phenomena is that it was a combination of factors including burnup, power history and temperature. The problem appears to be very sensitive to these factors, the interaction of which is not clearly understood. The model developed by Westinghouse provides a possible explanation but there is not sufficient data to establish confidence levels and sensitivity studies involving the key parameters have not been done. While several fixes to the problem have been discussed, no definitive fixes have been proposed. Without complete understanding of the phenomena, or fixes that clearly eliminate the problem the safety concern remains. The safety significance depends on the amount of shutdown margin lost due to incomplete insertion of the control rods. Were the control rods to stick high in the core, the reactor could not be shutdown by the control rods and other means such as emergency boration would be required.

  2. CONTROL ROD DRIVE

    DOEpatents

    Chapellier, R.A.

    1960-05-24

    BS>A drive mechanism was invented for the control rod of a nuclear reactor. Power is provided by an electric motor and an outside source of fluid pressure is utilized in conjunction with the fluid pressure within the reactor to balance the loadings on the motor. The force exerted on the drive mechanism in the direction of scramming the rod is derived from the reactor fluid pressure so that failure of the outside pressure source will cause prompt scramming of the rod.

  3. Pull rod assembly

    DOEpatents

    Cioletti, O.C.

    1988-04-21

    A pull rod assembly comprising a pull rod having three peripheral grooves, a piston device including an adaptor ring and a seal ring, said piston device being mounted on the pull rod by a split ring retainer situated in one groove and extending into an interior groove in the adaptor and a resilient split ring retained in another groove and positioned to engage the piston device and to retain the seal on its adaptor.

  4. HTGR green rod intercomparison. I. Hydrogen assay

    SciTech Connect

    Meier, M.M.; Adams, E.L.

    1981-01-01

    A /sup 252/Cf hydrogen monitor has been used to determine the hydrogen content of each of the 144 unfired (green) fuel rods being circulated in the New Brunswick Laboratory fuel measurement evaluation program. The monitor was calibrated with standards fabricated by Los Alamos Scientific Laboratory (LASL) Analytical Chemistry Group, CMB-1. Measurements were made relative to these standards and overall accuracies of +-3% were achieved.

  5. Rod internal pressure quantification and distribution analysis using Frapcon

    SciTech Connect

    Bratton, Ryan N; Jessee, Matthew Anderson; Wieselquist, William A

    2015-09-01

    This report documents work performed supporting the Department of Energy (DOE) Office of Nuclear Energy (NE) Fuel Cycle Technologies Used Fuel Disposition Campaign (UFDC) under work breakdown structure element 1.02.08.10, ST Analysis. In particular, this report fulfills the M4 milestone M4FT- 15OR0810036, Quantify effects of power uncertainty on fuel assembly characteristics, within work package FT-15OR081003 ST Analysis-ORNL. This research was also supported by the Consortium for Advanced Simulation of Light Water Reactors (http://www.casl.gov), an Energy Innovation Hub (http://www.energy.gov/hubs) for Modeling and Simulation of Nuclear Reactors under U.S. Department of Energy Contract No. DE-AC05-00OR22725. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each evaluated WBN1 fuel rod. An alternate model for the amount of helium released from the zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layer is derived and applied to FRAPCON output data to quantify the RIP and CHS for these types of fuel rods. SCALE/Polaris is used to quantify fuel rodspecific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel pellets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd

  6. Rod Photoreceptors Detect Rapid Flicker

    ERIC Educational Resources Information Center

    Conner, J. D.; MacLeod, Donald I. A.

    1977-01-01

    Rod-isolation techniques show that light-adapted human rods detect flicker frequencies as high as 28 hertz, and that the function relating rod critical flicker frequency to stimulus intensity contains two distinct branches. (MLH)

  7. Control rod driveline and grapple

    DOEpatents

    Germer, John H.

    1987-01-01

    A control rod driveline and grapple is disclosed for placement between a control rod drive and a nuclear reactor control rod containing poison for parasitic neutron absorption required for reactor shutdown. The control rod is provided with an enlarged cylindrical handle which terminates in an upwardly extending rod to provide a grapple point for the driveline. The grapple mechanism includes a tension rod which receives the upwardly extending handle and is provided with a lower annular flange. A plurality of preferably six grapple segments surround and grip the control rod handle. Each grapple rod segment grips the flange on the tension rod at an interior upper annular indentation, bears against the enlarged cylindrical handle at an intermediate annulus and captures the upwardly flaring frustum shaped handle at a lower and complementary female segment. The tension rods and grapple segments are surrounded by and encased within a cylinder. The cylinder terminates immediately and outward extending annulus at the lower portion of the grapple segments. Excursion of the tension rod relative to the encasing cylinder causes rod release at the handle by permitting the grapple segments to pivot outwardly and about the annulus on the tension rod so as to open the lower defined frustum shaped annulus and drop the rod. Relative movement between the tension rod and cylinder can occur either due to electromagnetic release of the tension rod within defined limits of travel or differential thermal expansion as between the tension rod and cylinder as where the reactor exceeds design thermal limits.

  8. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

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

    1963-06-11

    A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

  9. 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

  10. 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

  11. 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.

  12. The Self Actualized Reader.

    ERIC Educational Resources Information Center

    Marino, Michael; Moylan, Mary Elizabeth

    A study examined the commonalities that "voracious" readers share, and how their experiences can guide parents, teachers, and librarians in assisting children to become self-actualized readers. Subjects, 25 adults ranging in age from 20 to 67 years, completed a questionnaire concerning their reading histories and habits. Respondents varied in…

  13. 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

  14. 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

  15. 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

  16. Volume reduction of spent fuel elements for direct disposal

    SciTech Connect

    Wasserfuhr, I.C.

    1995-12-31

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

  17. Understanding flame rods

    SciTech Connect

    McAuley, J.A. Jr.

    1995-11-01

    The flame rod is probably the least understood method of flame detection. Although it is not recommended for oilfired equipment, it is very common on atmospheric, or {open_quotes}in-shot,{close_quotes} gas burners. It is also possible, although not common, to have an application with a constant gas pilot, monitored by a flame rod, and maintaining an oil main flame. Regardless of the application, chances are that flame rods will be encountered during the course of servicing. The technician today must be versatile and able to work on many different types of equipment. One must understand the basic principles of flame rods, and how to correct potential problems. The purpose of a flame detection system is two-fold: (1) to prove there is no flame when there shouldn`t be one, and (2) to prove there is a flame when there should be one. Flame failure response time is very important. This is the amount of time it takes to realize there is a loss of flame, two to four seconds is typical today. Prior to flame rods, either bi-metal or thermocouple type flame detectors were common. The response time for these detectors was up to three minutes, seldom less than one minute.

  18. Intramedullary rodding in osteogenesis imperfecta.

    PubMed

    Mulpuri, K; Joseph, B

    2000-01-01

    The results of intramedullary rodding of long bones of 16 children with osteogenesis imperfecta, over a 10-year period, were analyzed. Sheffield elongating rods or non-elongating rods were used. The frequency of fractures was dramatically reduced after implantation of either type of rod, and the ambulatory status improved in all instances. The results were significantly better after Sheffield rodding with regard to the frequency of complications requiring reoperations and the longevity of the rods. Migration of the rods, encountered frequently, appears to be related to improper placement of the rods in the bone. It seems likely that if care is taken to ensure precise placement of a rod of appropriate size, several of these complications may be avoided. PMID:10739296

  19. COMPOSITE CONTROL ROD

    DOEpatents

    Rock, H.R.

    1963-12-24

    A composite control rod for use in controlling a nuclear reactor is described. The control rod is of sandwich construction in which finned dowel pins are utilized to hold together sheets of the neutron absorbing material and nonabsorbing structural material thereby eliminating the need for being dependent on the absorbing material for structural support. The dowel pins perform the function of absorbing the forces due to differential thermal expansion, seating further with the fins into the sheets of material and crushing before damage is done either to the absorbing or non-absorbing material. (AEC)

  20. A thermodynamic model for noble metal alloy inclusions in nuclear fuel rods and application to the study of loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Kaye, Matthew Haigh

    Metal alloy inclusions comprised of Mo, Pd, Rh, Ru, and Tc (the so-called "noble" metals) develop in CANDU fuel pellets as a result of fission. The thermochemical behaviour of this alloy system during severe accident conditions is of interest in connection with computations of loss of volatile compounds of these elements by reaction with steam-hydrogen gas mixtures that develop in the system as a result of water reacting with the Zircalloy cladding. This treatment focuses on the development of thermodynamic models for the Mo-Pd-Rh-Ru-Tc quinary system. A reasonable prediction was made by modelling the ten binary phase diagrams, five of these evaluations being original to this work. This process provides a complete treatment for the five solution phases (vapour, liquid, bcc-solid, fcc-solid, and cph-solid) in this alloy system, as well as self-consistent Gibbs energies of formation for the Mo 5Ru3 intermetallic phase, and two intermediate phases in the Mo-Tc system. The resulting collection of properties, when treated by Gibbs energy minimization, permits phase equilibria to be computed for specified temperatures and compositions. Experimental work in support of this treatment has been performed. Measurements of the solidus and liquidus temperatures for Pd-Rh alloys were made using differential thermal analysis. These measurements confirm that the liquid solution exhibits positive deviation from Raoult's law. Experimental work as a visiting research engineer at AECL (Chalk River) was performed using a custom developed Knudsen cell/mass spectrometer. The Pd partial pressure was measured above multi-component alloys of known composition over a range of temperatures. These are correlated to predicted activities of Pd from the developed thermodynamic model in the multi-component alloy. The thermodynamic treatment developed for the noble metal alloy inclusions has been combined with considerable other data and applied to selected loss-of-coolant-accident scenarios to

  1. Coupled thermal analysis applied to the study of the rod ejection accident

    SciTech Connect

    Gonnet, M.

    2012-07-01

    An advanced methodology for the assessment of fuel-rod thermal margins under RIA conditions has been developed by AREVA NP SAS. With the emergence of RIA analytical criteria, the study of the Rod Ejection Accident (REA) would normally require the analysis of each fuel rod, slice by slice, over the whole core. Up to now the strategy used to overcome this difficulty has been to perform separate analyses of sampled fuel pins with conservative hypotheses for thermal properties and boundary conditions. In the advanced methodology, the evaluation model for the Rod Ejection Accident (REA) integrates the node average fuel and coolant properties calculation for neutron feedback purpose as well as the peak fuel and coolant time-dependent properties for criteria checking. The calculation grid for peak fuel and coolant properties can be specified from the assembly pitch down to the cell pitch. The comparative analysis of methodologies shows that coupled methodology allows reducing excessive conservatism of the uncoupled approach. (authors)

  2. Anchor for Fiberglas Guy Rod

    NASA Technical Reports Server (NTRS)

    Wilson, A. H.

    1982-01-01

    Solution to problem of anchoring fiberglas guy rods to install nut with threads on outer circumference, followed by aluminum sleeve. Sleeve has opening oval at upper and round at bottom end. End of rod is split so fiberglas wedge can be inserted to form V-shaped end. Spread end of rod fits into tapered hole in sleeve and threaded aluminum coupling is put over rod and sleeve.

  3. REACTOR CONTROL ROD OPERATING SYSTEM

    DOEpatents

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  4. Pattern fuel assembly loading system

    SciTech Connect

    Ahmed, H.J.; Gerkey, K.S.; Miller, T.W.; Wylie, M.E.

    1986-12-02

    This patent describes an interactive system for facilitating preloading of fuel rods into magazines, which comprises: an operator work station adapted for positioning between a supply of fuel rods of predetermined types, and the magazine defining grid locations for a predetermined fuel assembly; display means associated with the work station; scanner means associated with the work station and adapted for reading predetermined information accompanying the fuel rods; a rectangular frame adapted for attachment to one end of the fuel assembly loading magazine; prompter/detector means associated with the frame for detecting insertion of a fuel rod into the magazine; and processing means responsive to the scanner means and the sensing means for prompting the operator via the display means to pre-load the fuel rods into desired grid locations in the magazine. An apparatus is described for facilitating pre-loading of fuel rods in predetermined grid locations of a fuel assembly loading magazine, comprising: a rectangular frame adapted for attachment to one end of the fuel assembly loading magazine; and means associated with the frame for detecting insertion of fuel rods into the magazine.

  5. Hafnium stainless steel absorber rod for control rod

    SciTech Connect

    Charnley, J.E.; Cearley, J.E.; Dixon, R.C.; Izzo, K.R.; Aiello, L.L.

    1989-08-01

    This patent describes an improvement in a control rod having a stainless steel body for enclosing a neutron absorbing poison, the control rod having movement along an axial direction for insertion into and out of a nuclear reactor for controlling a nuclear reaction. The improvement comprising: a piece of hafnium; a piece of stainless steel joined to the hafnium by a thin diffusion interface created by friction welding. The hafnium and the stainless steel oriented serially in the axial direction with the thin diffusion interface disposed normal to the axial direction of the control rod movement; means for confining the hafnium to movement along the axial direction with the control rod; and means for attaching the piece of stainless steel to the remaining portion of the control rod to load the weld therebetween under compression or tension during the control rod movement. Whereby the thin diffusion interface is loaded in tension or compression only upon dynamic movement of the control rod.

  6. Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program.

    SciTech Connect

    Gregson, Michael Warren; Mo, Tin; Sorenson, Ken Bryce; Loiseau, Olivier; Nolte, Oliver; Hibbs, Russell S.; Molecke, Martin Alan; Slater-Thompson, Nancy; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Tsai, Han-Chung; Billone, Michael C.; Lange, Florentin; Young, Francis I.

    2004-08-01

    The authors provide a detailed overview of an on-going, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy-density device. The program participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research program. Sandia National Laboratories has the lead role for conducting this research program; test program support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. The authors provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. They focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests, and briefly summarize similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods.

  7. SAFETY SYSTEM FOR CONTROL ROD

    DOEpatents

    Paget, J.A.

    1963-05-14

    A structure for monitoring the structural continuity of a control rod foi a neutron reactor is presented. A electric conductor readily breakable under mechanical stress is fastened along the length of the control rod at a plurality of positions and forms a closed circuit with remote electrical components responsive to an open circuit. A portion of the conductor between the control rod and said components is helically wound to allow free and normally unrestricted movement of the segment of conductor secured to the control rod relative to the remote components. Any break in the circuit is indicative of control rod breakage. (AEC)

  8. Vortices in vibrated granular rods

    NASA Astrophysics Data System (ADS)

    Neicu, Toni; Kudrolli, Arshad

    2002-03-01

    We report the first experimental observation of vortex patterns in granular rods inside a container that is vibrated vertically . The experiments were carried out with an anodized aluminum circular container which is rigidly attached to an electromagnetic shaker and the patterns are imaged using a high-frame rate digital camera. At low rod numbers and driving amplitudes, the rods are observed to lie horizontally. Above a critical number or packing fraction of rods, moving domains of vertical rods are spontaneously observed to form which coexist with horizontal rods. These small domains of vertical rods coarsen over time to form a few large vortices. The size of the vortices increases with the number of rods. We are able to track the ends of the vertical rods and obtain the velocity fields of the vortices. The mean azimuthal velocity as a function of distance from the center of the vortex is obtained as a function of the packing fraction. We will report the phase diagram of the various patterns observed as function of number of rods and driving amplitude. The mechanism for the formation and motion of the domains of vertical rods will be also discussed.

  9. Safety rod latch inspection

    SciTech Connect

    Leader, D.R.

    1992-02-01

    During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small button'' in the latch mechanism had broken off of the lock plunger'' and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and metallography. During the examination of the L-Area latches, a failed latch mechanism from the 100 K reactor was added to the investigation. Fourteen latches that had a history of problems were removed from K-Area and sent to SRL for inclusion in this study the week after the original seven assemblies were examined, bringing the total of latch assemblies discussed in this report to twenty one. Results of the examination of the K-Area latch that initiated this study is not included in this report.

  10. Safety rod latch inspection

    SciTech Connect

    Leader, D.R.

    1992-02-01

    During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small ``button`` in the latch mechanism had broken off of the ``lock plunger`` and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and metallography. During the examination of the L-Area latches, a failed latch mechanism from the 100 K reactor was added to the investigation. Fourteen latches that had a history of problems were removed from K-Area and sent to SRL for inclusion in this study the week after the original seven assemblies were examined, bringing the total of latch assemblies discussed in this report to twenty one. Results of the examination of the K-Area latch that initiated this study is not included in this report.

  11. TEST SYSTEM FOR EVALUATING SPENT NUCLEAR FUEL BENDING STIFFNESS AND VIBRATION INTEGRITY

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom; Howard, Rob L; Flanagan, Michelle

    2013-01-01

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements specified by federal regulations. For normal conditions of transport, vibration loads incident to transport must be considered. This is particularly relevant for high-burnup fuel (>45 GWd/MTU). As the burnup of the fuel increases, a number of changes occur that may affect the performance of the fuel and cladding in storage and during transportation. The mechanical properties of high-burnup de-fueled cladding have been previously studied by subjecting defueled cladding tubes to longitudinal (axial) tensile tests, ring-stretch tests, ring-compression tests, and biaxial tube burst tests. The objective of this study is to investigate the mechanical properties and behavior of both the cladding and the fuel in it under vibration/cyclic loads similar to the sustained vibration loads experienced during normal transport. The vibration loads to SNF rods during transportation can be characterized by dynamic, cyclic, bending loads. The transient vibration signals in a specified transport environment can be analyzed, and frequency, amplitude and phase components can be identified. The methodology being implemented is a novel approach to study the vibration integrity of actual SNF rod segments through testing and evaluating the fatigue performance of SNF rods at defined frequencies. Oak Ridge National Laboratory (ORNL) has developed a bending fatigue system to evaluate the response of the SNF rods to vibration loads. A three-point deflection measurement technique using linear variable differential transformers is used to characterize the bending rod curvature, and electromagnetic force linear motors are used as the driving system for mechanical loading. ORNL plans to use the test system in a hot cell for SNF vibration testing on high burnup, irradiated fuel to evaluate the pellet-clad interaction and bonding on the effective lifetime of fuel-clad structure bending fatigue performance. Technical

  12. New results from the NSRR experiments with high burnup fuel

    SciTech Connect

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  13. Considerations for sensitivity analysis, uncertainty quantification, and data assimilation for grid-to-rod fretting

    SciTech Connect

    Michael Pernice

    2012-10-01

    Grid-to-rod fretting is the leading cause of fuel failures in pressurized water reactors, and is one of the challenge problems being addressed by the Consortium for Advanced Simulation of Light Water Reactors to guide its efforts to develop a virtual reactor environment. Prior and current efforts in modeling and simulation of grid-to-rod fretting are discussed. Sources of uncertainty in grid-to-rod fretting are also described.

  14. 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.

  15. Rod/Coil Block Copolyimides for Ion-Conducting Membranes

    NASA Technical Reports Server (NTRS)

    Meador, Mary Ann B.; Kinder, James D.

    2003-01-01

    Rod/coil block copolyimides that exhibit high levels of ionic conduction can be made into diverse products, including dimensionally stable solid electrolyte membranes that function well over wide temperature ranges in fuel cells and in lithium-ion electrochemical cells. These rod/coil block copolyimides were invented to overcome the limitations of polymers now used to make such membranes. They could also be useful in other electrochemical and perhaps some optical applications, as described below. The membranes of amorphous polyethylene oxide (PEO) now used in lithium-ion cells have acceptably large ionic conductivities only at temperatures above 60 C, precluding use in what would otherwise be many potential applications at lower temperatures. PEO is difficult to process, and, except at the highest molecular weights it is not very dimensionally stable. It would be desirable to operate fuel cells at temperatures above 80 C to take advantage of better kinetics of redox reactions and to reduce contamination of catalysts. Unfortunately, proton-conduction performance of a typical perfluorosulfonic polymer membrane now used as a solid electrolyte in a fuel cell decreases with increasing temperature above 80 C because of loss of water from within the membrane. The loss of water has been attributed to the hydrophobic nature of the polymer backbone. In addition, perfluorosulfonic polymers are expensive and are not sufficiently stable for long-term use. Rod/coil block copolyimides are so named because each molecule of such a polymer comprises short polyimide rod segments alternating with flexible polyether coil segments (see figure). The rods and coils can be linear, branched, or mixtures of linear and branched. A unique feature of these polymers is that the rods and coils are highly incompatible, giving rise to a phase separation with a high degree of ordering that creates nanoscale channels in which ions can travel freely. The conduction of ions can occur in the coil phase

  16. Fiber optic laser rod

    DOEpatents

    Erickson, G.F.

    1988-04-13

    A laser rod is formed from a plurality of optical fibers, each forming an individual laser. Synchronization of the individual fiber lasers is obtained by evanescent wave coupling between adjacent optical fiber cores. The fiber cores are dye-doped and spaced at a distance appropriate for evanescent wave coupling at the wavelength of the selected dye. An interstitial material having an index of refraction lower than that of the fiber core provides the optical isolation for effective lasing action while maintaining the cores at the appropriate coupling distance. 2 figs.

  17. Making Highly Pure Glass Rods

    NASA Technical Reports Server (NTRS)

    Naumann, R. J.

    1986-01-01

    Proposed quasi-containerless method for making glass rods or fibers minimizes contact between processing equipment and product. Method allows greater range of product sizes and shapes than achieved in experiments on containerless processing. Molten zone established in polycrystalline rod. Furnace sections separated, and glass rod solidifies between them. Clamp supports solid glass as it grows in length. Pulling clamp rapidly away from melt draws glass fiber. Fiber diameter controlled by adjustment of pulling rate.

  18. Automatic safety rod for reactors

    DOEpatents

    Germer, John H.

    1988-01-01

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  19. Piston and connecting rod assembly

    NASA Technical Reports Server (NTRS)

    Brogdon, James William (Inventor); Gill, David Keith (Inventor); Chatten, John K. (Inventor)

    2001-01-01

    A piston and connecting rod assembly includes a piston crown, a piston skirt, a connecting rod, and a bearing insert. The piston skirt is a component separate from the piston crown and is connected to the piston crown to provide a piston body. The bearing insert is a component separate from the piston crown and the piston skirt and is fixedly disposed within the piston body. A bearing surface of a connecting rod contacts the bearing insert to thereby movably associate the connecting rod and the piston body.

  20. Determination of the bias in LOFT fuel peak cladding temperature data from the blowdown phase of large-break LOCA experiments

    SciTech Connect

    Berta, V.T.; Hanson, R.G.; Johnsen, G.W.; Schultz, R.R.

    1993-05-01

    Data from the Loss-of-Fluid Test (LOFT) Program help quantify the margin of safety inherent in pressurized water reactors during postulated loss-of-coolant accidents (LOCAs). As early as 1979, questions arose concerning the accuracy of LOFT fuel rod cladding temperature data during several large-break LOCA experiments. This report analyzes how well externally-mounted fuel rod cladding thermocouples in LOFT accurately reflected actual cladding surface temperature during large-break LOCA experiments. In particular, the validity of the apparent core-wide fuel rod cladding quench exhibited during blowdown in LOFT Experiments L2-2 and L2-3 is studied. Also addressed is the question of whether the externally-mounted thermocouples might have influenced cladding temperature. The analysis makes use of data and information from several sources, including later, similar LOFT Experiments in which fuel centerline temperature measurements were made, experiments in other facilities, and results from a detailed FRAP-T6 model of the LOFT fuel rod. The analysis shows that there can be a significant difference (referred to as bias) between the surface-mounted thermocouple reading and the actual cladding temperature, and that the magnitude of this bias depends on the rate of heat transfer between the fuel rod cladding and coolant. The results of the analysis demonstrate clearly that a core-wide cladding quench did occur in Experiments L2-2 and L2-3. Further, it is shown that, in terms of peak cladding temperature recording during LOFT large-break LOCA experiments, the mean bias is 11.4 {plus_minus} 16.2K (20.5 {plus_minus} 29.2{degrees} F). The best-estimate value of peak cladding temperature for LOFT LP-02-6 is 1,104.8 K. The best-estimate peak cladding temperature for LOFT LP-LB-1 is 1284.0 K.

  1. Materials and mechanical design analysis of boron carbide reactor safety rods. Final report

    SciTech Connect

    Marra, J.C.

    1992-04-01

    The purpose of this task was to analyze the materials and mechanical design bases for the new boron carbide safety rod. These analyses included examination of the irradiation response of the materials, chemical compatibility of component materials, moisture considerations for the boron carbide pellets and susceptibility of the rod to corrosion under reactor environmental conditions. A number of issues concerning the mechanical behavior were also addressed. These included: safety rod dynamic response in scram scenarios, flexibility and mishandling behavior, and response to thermal excursions associated with gamma heating. A surveillance program aimed at evaluating the integrity of the safety rods following actual operating conditions and justifying life extension for the rods was also proposed. Based on the experimental testing and analyses associated with this task, it is concluded that the boron carbide safety rod design meets the materials and mechanical criteria for successful operational performance.

  2. Spherical Joint Piston and Connecting Rod Developed

    NASA Technical Reports Server (NTRS)

    1996-01-01

    Under an interagency agreement with the Department of Energy, the NASA Lewis Research Center manages a Heavy-Duty Diesel Engine Technology (HDET) research program. The overall program objectives are to reduce fuel consumption through increased engine efficiency, reduce engine exhaust emissions, and provide options for the use of alternative fuels. The program is administered with a balance of research contracts, university research grants, and focused in-house research. The Cummins Engine Company participates in the HDET program under a cost-sharing research contract. Cummins is researching and developing in-cylinder component technologies for heavy-duty diesel engines. An objective of the Cummins research is to develop technologies for a low-emissions, 55-percent thermal efficiency (LE-55) engine. The best current-production engines in this class achieve about 46-percent thermal efficiency. Federal emissions regulations are driving this technology. Regulations for heavy duty diesel engines were tightened in 1994, more demanding emissions regulations are scheduled for 1998, and another step is planned for 2002. The LE-55 engine emissions goal is set at half of the 1998 regulation level and is consistent with plans for 2002 emissions regulations. LE-55 engine design requirements to meet the efficiency target dictate a need to operate at higher peak cylinder pressures. A key technology being developed and evaluated under the Cummins Engine Company LE-55 engine concept is the spherical joint piston and connecting rod. Unlike conventional piston and connecting rod arrangements which are joined by a pin forming a hinged joint, the spherical joint piston and connecting rod use a ball-and-socket joint. The ball-and-socket arrangement enables the piston to have an axisymmetric design allowing rotation within the cylinder. The potential benefits of piston symmetry and rotation are reduced scuffing, improved piston ring sealing, improved lubrication, mechanical and thermal

  3. Composite Lightning Rods for Aircraft

    NASA Technical Reports Server (NTRS)

    Bryan, Charles F., Jr.

    1986-01-01

    Composite, lightweight sacrificial tip with graphite designed reduces lightning-strike damage to composite parts of aircraft and dissipates harmful electrical energy. Device consists of slender composite rod fabricated from highly-conductive unidirectional reinforcing fibers in matrix material. Rods strategically installed in trailing edges of aircraft wings, tails, winglets, control surfaces, and rearward-most portion of aft fuselage.

  4. Rod Climbing of Suspensions

    NASA Astrophysics Data System (ADS)

    Guo, Youjing; Wang, Xiaorong

    We wish to report an unexpected effect observed for particle suspensions sucked to pass through a vertical pipe. Above a critical concentration, the suspension on the outside of the pipe may climb along the outside wall of the pipe and then display a surprising rod-climbing effect. Our study shows that the phenomenon is influenced mainly by the suspension composition, the pipe dimension and the suction speed. The effects of the pipe materials of different kinds are negligible. Increasing the suction force and the concentration increases the climbing height. Increasing the pipe diameter and wall thickness reduces the climbing effect. This behavior may be relevant to that the suspensions of the type described are all displaying markedly shear-thickening.

  5. FUEL ELEMENT

    DOEpatents

    Bean, R.W.

    1963-11-19

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

  6. Shippingport LWBR (Th/U Oxide) Fuel Characteristics for Disposal Criticality Analysis

    SciTech Connect

    L. L. Taylor; H. H. Loo

    1999-09-01

    Department of Energy (DOE)-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into eight characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. The Shippingport Light Water Breeder Reactor (LWBR) fuels incorporate more of the conventional materials (zirconium cladding/heavy metal oxides) and fabrication details (rods and spacers) that make them comparable to a typical commercial fuel assembly. The LWBR seed/blanket configuration tested a light-water breeder concept with Th-232/U-233 binary fuel matrix. Reactor design used several assembly configurations at different locations within the same core . The seed assemblies contain the greatest fissile mass per (displaced) unit volume, but the blanket assemblies actually contain more fissile mass in a larger volume; the atom-densities are comparable.

  7. HIDUTYDRV Code, A Fuel Product Margin Tool

    SciTech Connect

    Krammen, Michael A.; Karoutas, Zeses E.; Grill, Steven F.; Sutharshan, Balendra

    2007-07-01

    HIDUTYDRV is a computer code currently used in core design to model the best estimate steady-state fuel rod corrosion performance for Westinghouse's CE-design 14x14 and 16x16 fuel. The fuel rod oxide thickness, sub-cooled nucleate boiling (referred to as mass evaporation or steaming), and fuel duty indices can be predicted for individual rods or up to every fuel rod in the quarter core at every nuclear fuel management depletion time-step as a function of axial elevation within the core. Best estimate operating margins for fuel components whose performance depends on the local power and thermal hydraulic conditions are candidates for analysis with HIDUTYDRV. HIDUTYDRV development will focus on fuel component parameters associated with known leakers for addressing INPO goals to eliminate fuel leakers by 2010. (authors)

  8. Technical description of the NRC long-term whole-rod and crud performance test

    SciTech Connect

    Einziger, R.E.; Fish, R.L.; Knecht, R.L.

    1982-09-01

    Westinghouse Hanford Company (WHC) and EG and G-Idaho are jointly conducting a long-term, low-temperature, spent-fuel, whole rod and crud behavior test to provide the Nuclear Regulatory Commission (NRC) with information to assist in the licensing of light water reactor (LWR) spent-fuel, dry storage facilities. Readily available fuel rods from an H.B. Robinson Unit 2 (PWR) fuel assembly and a Peach Bottom-II (BWR) fuel assembly were selected for use in the 50-month test. Both intact and defected rods will be tested in inert and oxidizing atmospheres. A 230/sup 0/C test temperature was selected for the first 10-month run. Both nondestructive and destructive examinations are planned to characterize the fuel rod behavior during the 5-y test. Four interim examinations and a final examination will be conducted. Crud spallation behavior will be investigated by sampling the crud particulate from the test capsules at each of the four interim examinations and at the end of the test. The background to whole rod testing, description of rod breach mechanisms, and a detailed description of the test are presented in this document.

  9. Regenerative hyperpolarization in rods.

    PubMed Central

    Werblin, F S

    1975-01-01

    1. The electrical properties of the rods in Necturus maculosus were studied at the cell body and the outer segments in dark and light under current and voltage clamp with a pair of intracellular electrodes separated by about 1 mum. 2. The membrane resistance in the dark was voltage- and time-dependent both for the cell body and the outer segment. Slight depolarizations in the cell body reduced the slope resistance from 60 to 10 M omega with a time constant of about 1 sec. Polarization in either direction, at the outer segment, when greater than about 20 mV, reduced the slope resistance from 60 to 30 M omega. The dark potential in the cell body was typically -30 to -35 m V; at the outer segment it was typically only -10 to -15 mV. 3. The light-elicited voltage response in both the cell body and the outer segment was largest with the membrane near the dark potential level. In both regions, the response was reduced when the membrane was polarized in either direction. 4. Under voltage-clamp conditions, a reversal potential for the light response near + 10 mV was measured at the outer segment. At the cell body no reversal potential for the light response was measured; there the clamping current required during the light response was almost of the same magnitude at all potential levels. 5. When the membrane at the cell body was hyperpolarized in the dark under voltage clamp, a transient outward current, typically about one-half the magnitude of the initial inward clamping current was required to maintain the membrane at the clamped potential level. This outward current transient was associated with a decrease in membrane resistance with similar time course. The transient outward current reversed and became inward when the membrane was clamped to potentials more negative than -80 mV. Thus, the transient outward current appears to involve a transient activation initiated by hyperpolarization. I is regenerative in that it is initiated by hyperpolarization and tends to

  10. Hopf solitons and elastic rods

    SciTech Connect

    Harland, Derek; Sutcliffe, Paul; Speight, Martin

    2011-03-15

    Hopf solitons in the Skyrme-Faddeev model are stringlike topological solitons classified by the integer-valued Hopf charge. In this paper we introduce an approximate description of Hopf solitons in terms of elastic rods. The general form of the elastic rod energy is derived from the field theory energy and is found to be an extension of the classical Kirchhoff rod energy. Using a minimal extension of the Kirchhoff energy, it is shown that a simple elastic rod model can reproduce many of the qualitative features of Hopf solitons in the Skyrme-Faddeev model. Features that are captured by the model include the buckling of the charge three solution, the formation of links at charges five and six, and the minimal energy trefoil knot at charge seven.

  11. 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.

  12. NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

    DOEpatents

    Oakes, L.C.; Walker, C.S.

    1959-12-15

    ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

  13. Thermal breeder fuel enrichment zoning

    DOEpatents

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

    1992-01-01

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

  14. Support grid for fuel elements in a nuclear reactor

    DOEpatents

    Finch, Lester M.

    1977-01-01

    A support grid is provided for holding nuclear fuel rods in a rectangular array. Intersecting sheet metal strips are interconnected using opposing slots in the strips to form a rectangular cellular grid structure for engaging the sides of a multiplicity of fuel rods. Spring and dimple supports for engaging fuel and guide rods extending through each cell in the support grid are formed in the metal strips with the springs thus formed being characterized by nonlinear spring rates.

  15. 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.

  16. Noise reduction in a Mach 5 wind tunnel with a rectangular rod-wall sound shield

    NASA Technical Reports Server (NTRS)

    Creel, T. R., Jr.; Keyes, J. W.; Beckwith, I. E.

    1980-01-01

    A rod wall sound shield was tested over a range of Reynolds numbers of 0.5 x 10 to the 7th power to 8.0 x 10 to the 7th power per meter. The model consisted of a rectangular array of longitudinal rods with boundary-layer suction through gaps between the rods. Suitable measurement techniques were used to determine properties of the flow and acoustic disturbance in the shield and transition in the rod boundary layers. Measurements indicated that for a Reynolds number of 1.5 x 10 to the 9th power the noise in the shielded region was significantly reduced, but only when the flow is mostly laminar on the rods. Actual nozzle input noise measured on the nozzle centerline before reflection at the shield walls was attenuated only slightly even when the rod boundary layer were laminar. At a lower Reynolds number, nozzle input noise at noise levels in the shield were still too high for application to a quiet tunnel. At Reynolds numbers above 2.0 x 10 the the 7th power per meter, measured noise levels were generally higher than nozzle input levels, probably due to transition in the rod boundary layers. The small attenuation of nozzle input noise at intermediate Reynolds numbers for laminar rod layers at the acoustic origins is apparently due to high frequencies of noise.

  17. Advanced gray rod control assembly

    DOEpatents

    Drudy, Keith J; Carlson, William R; Conner, Michael E; Goldenfield, Mark; Hone, Michael J; Long, Jr., Carroll J; Parkinson, Jerod; Pomirleanu, Radu O

    2013-09-17

    An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber to enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.

  18. Linguistic Theory and Actual Language.

    ERIC Educational Resources Information Center

    Segerdahl, Par

    1995-01-01

    Examines Noam Chomsky's (1957) discussion of "grammaticalness" and the role of linguistics in the "correct" way of speaking and writing. It is argued that the concern of linguistics with the tools of grammar has resulted in confusion, with the tools becoming mixed up with the actual language, thereby becoming the central element in a metaphysical…

  19. CRC DEPLETION CALCULATIONS FOR THE RODDED ASSEMBLIES IN BATCHES 1, 2, 3, AND 1X OF CRYSTAL RIVER UNIT 3

    SciTech Connect

    Kenneth D. Wright

    1997-09-03

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain rodded fuel assemblies from batches 1, 2, 3, and 1X of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A rodded assembly is one that contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) for some period of time during its irradiation history. The objective of this analysis is to provide SAS2H calculated isotopic compositions of depleted fuel and depleted burnable poison for each fuel assembly to be used in subsequent CRC reactivity calculations containing the fuel assemblies.

  20. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

    SciTech Connect

    Kenneth D. Wright

    1997-07-30

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  1. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 1, 2, AND 3 OF CRYSTAL RIVER UNIT 3

    SciTech Connect

    Kenneth D. Wright

    1997-07-29

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 1, 2, and 3 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  2. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 8 AND 9 CRYSTAL RIVER UNIT 3

    SciTech Connect

    Michael L. Wilson

    2001-02-08

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 8 and 9 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  3. 40 CFR 74.22 - Actual SO2 emissions rate.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ....6 for natural gas For other fuels, the combustion source must specify the SO2 emissions factor. (c... (CONTINUED) SULFUR DIOXIDE OPT-INS Allowance Calculations for Combustion Sources § 74.22 Actual SO2 emissions rate. (a) Data requirements. The designated representative of a combustion source shall submit...

  4. Fuel behavior during a LOCA: LOFT experiments

    SciTech Connect

    Russell, M.L.

    1980-11-01

    The LOFT experiments have provided the following fuel behavior information which appears to be valuable for improving the safety of PWR operation and resolving PWR licensing issues: (1) A generic unassisted core cooling event occurs during large-break LOCAs that dominates the cooling of the core before ECC reflood commences and potentially eliminates the possibility of flow channel blockage from prepressurized fuel rod swelling. (2) The large-break LOCA decompression forces do not disturb the normal control rod gravity drop and may not structually damage the fuel assemblies. (3) Large-break LOCA core cooling may also be enhanced by spacer grid and core counter flow delay of liquid escape from the core boundaries and liquid fallback from the upper plenum into the core region. (4) Lower fuel rod prepressurization may be possible in PWR fuel rods which would reduce flow channel blockage complications during LOCA's. (5) Uniform fuel rod cladding temperature indications during the large break LOCA's do not confirm expectations for the fuel rod cladding temperature variations that would inhibit development of flow channel blockages by ballooning of prepressurized fuel rods.

  5. COST IMPACT OF ROD CONSOLIDATION ON THE VIABILITY ASSESSMENT DESIGN

    SciTech Connect

    D. Lancaster

    1999-03-29

    The cost impact to the Civilian Radioactive Waste Management System of using rod consolidation is evaluated. Previous work has demonstrated that the fuel rods of two assemblies can be packed into a canister that can fit into the same size space as that used to store a single assembly. The remaining fuel assembly hardware can be compacted into the same size canisters with a ratio of 1 hardware canister per each 6 to 12 assemblies. Transportation casks of the same size as currently available can load twice the number of assemblies by placing the compacted assemblies in the slots currently designed for a single assembly. Waste packages similarly could contain twice the number of assemblies; however, thermal constraints would require considering either a low burnup or cooling. The analysis evaluates the impact of rod consolidation on CRWMS costs for consolidation at prior to transportation and for consolidation at the Monitored Geological Repository surface facility. For this study, no design changes were made to either the transport casks or waste packages. Waste package designs used for the Viability Assessment design were employed but derated to make the thermal limits. A logistics analysis of the waste was performed to determine the number of each waste package with each loading. A review of past rod consolidation experience found cost estimates which range from $10/kgU to $32/kgU. $30/kgU was assumed for rod consolidation costs prior to transportation. Transportation cost savings are about $17/kgU and waste package cost savings are about $21/kgU. The net saving to the system is approximately $500 million if the consolidation is performed prior to transportation. If consolidation were performed at the repository surface facilities, it would cost approximately $15/kgU. No transportation savings would be realized. The net savings for consolidation at the repository site would be about $400 million dollars.

  6. Reconstitutable control assembly having removable control rods with detachable split upper end plugs

    SciTech Connect

    Gjertsen, R.K.; Knott, R.P.; Sparrow, J.A.

    1989-12-19

    This patent describes, in a reconstitutable control assembly for use with a nuclear fuel assembly, the control assembly including a spider structure and at least one control rod, an attachment joint for detachable fastening the control rod to the spider structure. The attachment joint comprising: a hollow connecting finger on the spider structure; and an elongated detachable split upper end plug on the control rod having a pair of separate upper and lower plug portions, the upper plug portion having integrally-connected tandemly- arranged upper, middle and lower sections. The lower plug portion having integrally-connected tandemly-arranged upper, middle and lower segments.

  7. Used Fuel Testing Transportation Model

    SciTech Connect

    Ross, Steven B.; Best, Ralph E.; Maheras, Steven J.; Jensen, Philip J.; England, Jeffery L.; LeDuc, Dan

    2014-09-24

    This report identifies shipping packages/casks that might be used by the Used Nuclear Fuel Disposition Campaign Program (UFDC) to ship fuel rods and pieces of fuel rods taken from high-burnup used nuclear fuel (UNF) assemblies to and between research facilities for purposes of evaluation and testing. Also identified are the actions that would need to be taken, if any, to obtain U.S. Nuclear Regulatory (NRC) or other regulatory authority approval to use each of the packages and/or shipping casks for this purpose.

  8. Three-Rod Linear Ion Traps

    NASA Technical Reports Server (NTRS)

    Janik, Gary R.; Prestage, John D.; Maleki, Lutfollah

    1993-01-01

    Three-parallel-rod electrode structures proposed for use in linear ion traps and possibly for electrostatic levitation of macroscopic particles. Provides wider viewing angle because they confine ions in regions outside rod-electrode structures.

  9. Control rods in LMFBRs: a physics assessment

    SciTech Connect

    McFarlane, H.F.; Collins, P.J.

    1982-08-01

    This physics assessment is based on roughly 300 control rod worth measurements in ZPPR from 1972 to 1981. All ZPPR assemblies simulated mixed-oxide LMFBRs, representing sizes of 350, 700, and 900 MWe. Control rod worth measurements included single rods, various combinations of rods, and Ta and Eu rods. Additional measurements studied variations in B/sub 4/C enrichment, rod interaction effects, variations in rod geometry, neutron streaming in sodium-filled channels, and axial worth profiles. Analyses were done with design-equivalent methods, using ENDF/B Version IV data. Some computations for the sensitivities to approximations in the methods have been included. Comparisons of these analyses with the experiments have allowed the status of control rod physics in the US to be clearly defined.

  10. Solid-state-laser-rod holder

    DOEpatents

    Gettemy, D.J.; Barnes, N.P.; Griggs, J.E.

    1981-08-11

    The disclosure relates to a solid state laser rod holder comprising Invar, copper tubing, and epoxy joints. Materials and coefficients of expansion of the components of the holder combine with the rod to produce a joint which will give before the rod itself will. The rod may be lased at about 70 to 80/sup 0/K and returned from such a temperature to room temperature repeatedly without its or the holder's destruction.

  11. 21 CFR 876.4270 - Colostomy rod.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Colostomy rod. 876.4270 Section 876.4270 Food and... GASTROENTEROLOGY-UROLOGY DEVICES Surgical Devices § 876.4270 Colostomy rod. (a) Identification. A colostomy rod is a device used during the loop colostomy procedure. A loop of colon is surgically brought out...

  12. 21 CFR 876.4270 - Colostomy rod.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 8 2014-04-01 2014-04-01 false Colostomy rod. 876.4270 Section 876.4270 Food and... GASTROENTEROLOGY-UROLOGY DEVICES Surgical Devices § 876.4270 Colostomy rod. (a) Identification. A colostomy rod is a device used during the loop colostomy procedure. A loop of colon is surgically brought out...

  13. 21 CFR 876.4270 - Colostomy rod.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 8 2013-04-01 2013-04-01 false Colostomy rod. 876.4270 Section 876.4270 Food and... GASTROENTEROLOGY-UROLOGY DEVICES Surgical Devices § 876.4270 Colostomy rod. (a) Identification. A colostomy rod is a device used during the loop colostomy procedure. A loop of colon is surgically brought out...

  14. 21 CFR 876.4270 - Colostomy rod.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 8 2012-04-01 2012-04-01 false Colostomy rod. 876.4270 Section 876.4270 Food and... GASTROENTEROLOGY-UROLOGY DEVICES Surgical Devices § 876.4270 Colostomy rod. (a) Identification. A colostomy rod is a device used during the loop colostomy procedure. A loop of colon is surgically brought out...

  15. Inverted Control Rod Lock-In Device

    DOEpatents

    Brussalis, W. G.; Bost, G. E.

    1962-12-01

    A mechanism which prevents control rods from dropping out of the reactor core in the event the vessel in which the reactor is mounted should capsize is described. The mechanism includes a pivoted toothed armature which engages the threaded control rod lead screw and prevents removal of the rod whenever the armature is not attracted by the provided electromagnetic means. (AEC)

  16. 21 CFR 876.4270 - Colostomy rod.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Colostomy rod. 876.4270 Section 876.4270 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES GASTROENTEROLOGY-UROLOGY DEVICES Surgical Devices § 876.4270 Colostomy rod. (a) Identification. A colostomy rod...

  17. Assessment of spent-fuel waste-form/stabilizer alternatives for geologic disposal

    NASA Astrophysics Data System (ADS)

    Einziger, R. E.; Himes, D. A.

    1982-09-01

    The possibility of burying canisterized unreprocessed spent fuel in a deep geologic repository is studied. One aspect is an assessment of the possible spent fuel waste forms. The fuel performance portion was to evaluate five candidate spent fuel waste forms for postemplacement performance with emphasis on their ability to retard the release of radiouclides to the repository geology. Spent fuel waste forms under general consideration were: (1) unaltered fuel assembly; (2) fuel assembly with end fittings removed to shorten the length; (3) rods vented to remove gases and resealed; (4) disassembled fuel bundles to close pack the rods; and (5) rods chopped and fragments immobilized in a matrix material.

  18. Synaptic transmission from rods to rod-dominated bipolar cells in the tiger salamander retina.

    PubMed

    Yang, X L; Wu, S M

    1993-06-11

    Synaptic transmission between photoreceptors and bipolar cells was studied in dark-adapted tiger salamander retinas. Based on the relative light sensitivity, bipolar cells, either depolarizing (DBC) or hyperpolarizing (HBC), fell into two groups: one receives inputs primarily from rods (rod-dominated bipolar cells, DBCR and HBCR) and the other receives inputs primarily from cones (cone-dominated bipolar cells, DBCC and HBCC). The input-output relations of the rod-DBCR and rod-HBCR synapses were determined by plotting the voltage responses of the rod and DBCR (or HBCR) to dim 500-nm light steps, which polarizes only the rods but not the cones. The slope gains of both synapses were the highest near the dark rod voltage (-2.5 for the rod-DBCR synapse and 4.0 for the rod-HBCR synapse), and they (the absolute values) became progressively smaller at more hyperpolarized rod voltages. PMID:8186975

  19. Determination of plutonium content in TRR spent fuel by nondestructive neutron counting

    NASA Astrophysics Data System (ADS)

    Chen, Yen-Fu; Sheu, Rong-Jiun; Chiao, Ling-Huan; Yuan, Ming-Chen; Jiang, Shiang-Huei

    2010-07-01

    For the nuclear safeguard purpose, this work aims to nondestructively determine the plutonium content in the Taiwan Research Reactor (TRR) spent fuel rods in the storage pool before the stabilization process, which transforms the metal spent fuel rods into oxide powder. A SPent-fuel-Neutron-Counter (SPNC) system was designed and constructed to carry out underwater scan measurements of neutrons emitting from the spent fuel rod, from which the 240Pu mass in the fuel rod will be determined. The SAS2 H control module of the SCALE 5.1 code package was applied to calculate the 240Pu-to-Pu mass ratio in the TRR spent fuel rod according to the given power history. This paper presents the methodology and design of our detector system as well as the measurements of four TRR spent fuel rods in the storage pool and the comparison of the measured results with the facility declared values.

  20. How People Actually Use Thermostats

    SciTech Connect

    Meier, Alan; Aragon, Cecilia; Hurwitz, Becky; Mujumdar, Dhawal; Peffer, Therese; Perry, Daniel; Pritoni, Marco

    2010-08-15

    Residential thermostats have been a key element in controlling heating and cooling systems for over sixty years. However, today's modern programmable thermostats (PTs) are complicated and difficult for users to understand, leading to errors in operation and wasted energy. Four separate tests of usability were conducted in preparation for a larger study. These tests included personal interviews, an on-line survey, photographing actual thermostat settings, and measurements of ability to accomplish four tasks related to effective use of a PT. The interviews revealed that many occupants used the PT as an on-off switch and most demonstrated little knowledge of how to operate it. The on-line survey found that 89% of the respondents rarely or never used the PT to set a weekday or weekend program. The photographic survey (in low income homes) found that only 30% of the PTs were actually programmed. In the usability test, we found that we could quantify the difference in usability of two PTs as measured in time to accomplish tasks. Users accomplished the tasks in consistently shorter times with the touchscreen unit than with buttons. None of these studies are representative of the entire population of users but, together, they illustrate the importance of improving user interfaces in PTs.

  1. ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

    DOEpatents

    Young, J.N.

    1958-04-22

    An electromagnetic apparatus for moving a rod-like member in small steps in either direction is described. The invention has particular application in the reactor field where the reactor control rods must be moved only a small distance and where the use of mechanical couplings is impractical due to the high- pressure seals required. A neutron-absorbing rod is mounted in a housing with gripping uaits that engage the rod, and coils for magnetizing the gripping units to make them grip, shift, and release the rod are located outside the housing.

  2. Rod cluster having improved vane configuration

    SciTech Connect

    Shockling, L.A.; Francis, T.A.

    1989-09-05

    This patent describes a pressurized water reactor vessel, the vessel defining a predetermined axial direction of the flow of coolant therewithin and having plural spider assemblies supporting, for vertical movement within the vessel, respective clusters of rods in spaced, parallel axial relationship, parallel to the predetermined axial direction of coolant flow, and a rod guide for each spider assembly and respective cluster of rods. The rod guide having horizontally oriented support plates therewithin, each plate having an interior opening for accommodating axial movement therethrough of the spider assembly and respective cluster of rods. The opening defining plural radially extending channels and corresponding parallel interior wall surfaces of the support plate.

  3. Exploiting rod technology. Final report

    SciTech Connect

    1990-06-01

    ROD development was proceeding apace until recent budgetary decisions caused funding support for ROD development to be drastically reduced. The funding which was originally provided by DARPA and the Balanced Technology Initiative (BTI) Office has been cut back to zero from $800K. To determine the aeroballistic coefficients of a candidate dart, ARDEC is currently supporting development out of its own 6.2 funds at about $100K. ARDEC has made slow progress toward achieving this end because of failures in the original dart during testing. It appears that the next dart design to be tested will diverge from the original concept visualized by DARPA and Science and Technology Associates (STA). STA, the design engineer, takes exception to these changes on the basis of inappropriate test conditions and insufficient testing. At this time, the full resolution of this issue will be difficult because of the current management structure, which separates the developer (ARDEC) from the designer (STA).

  4. Shielding and Containment Evaluations of the NAC-LWT Cask with Tritium Burnable Poison Rods

    SciTech Connect

    Holger Pfeifer; Norman Meinert

    2000-06-04

    In 1989, the NAC legal weight truck cask (NAC-LWT) was approved by the U.S. Nuclear Regulatory Commission to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Since that time, license amendments have allowed the shipment of high-burnup PWR and BWR fuel rods, MTR-type research reactor fuel elements, and TRIGA-type fuel elements. In 1999, DOE approved an NAC-LWT submittal for a shipment of lead test assemblies (LTAs) containing tritium-producing burnable poison rods (TPBARs). This paper presents the 10 CFR Part 71 shielding and containment evaluations of the NAC-LWT with the LTA payload.

  5. Shielding and containment evaluations of the NAC-LWT cask with tritium burnable poison rods

    SciTech Connect

    Pfeifer, H.; Meinert, N.

    2000-07-01

    In 1989, the NAC legal weight truck cask (NAC-LWT) was approved by the US Nuclear Regulatory Commission to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Since that time, license amendments have allowed the shipment of high-burnup PWR and BWR fuel rods, MTR-type research reactor fuel elements, and TRIGA-type fuel elements. In 1999, DOE approved an NAC-LWT submittal for the shipment of lead test assemblies (LTAs) containing tritium-producing burnable poison rods (TPBARs). This paper presents the 10 CFR Part 71 shielding and containment evaluations of the NAC-LWT with the LTA payload.

  6. 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.

  7. 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.

  8. Fission product release from irradiated LWR fuel under accident conditions

    SciTech Connect

    Strain, R.V.; Sanecki, J.E.; Osborne, M.F.

    1984-01-01

    Fission product release from irradiated LWR fuel is being studied by heating fuel rod segments in flowing steam and an inert carrier gas to simulate accident conditions. Fuels with a range of irradiation histories are being subjected to several steam flow rates over a wide range of temperatures. Fission product release during each test is measured by gamma spectroscopy and by detailed examination of the collection apparatus after the test has been completed. These release results are complemented by a detailed posttest examination of samples of the fuel rod segment. Results of release measurements and fuel rod characterizations for tests at 1400 through 2000/sup 0/C are presented in this paper.

  9. Reversible Bending Fatigue Testing on Zry-4 Surrogate Rods

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom; Howard, Rob L

    2014-01-01

    Testing high-burnup spent nuclear fuel (SNF) presents many challenges in areas such as specimen preparation, specimen installation, mechanical loading, load control, measurements, data acquisition, and specimen disposal because these tasks are complicated by the radioactivity of the test specimens. Research and comparison studies conducted at Oak Ridge National Laboratory (ORNL) resulted in a new concept in 2010 for a U-frame testing setup on which to perform hot-cell reversible bending fatigue testing. Subsequently, the three-dimensional finite element analysis and the engineering design of components were completed. In 2013 the ORNL team finalized the upgrade of the U-frame testing setup and the integration of the U-frame setup into a Bose dual linear motor test bench to develop a cyclic integrated reversible-bending fatigue tester (CIRFT). A final check was conducted on the CIRFT test system in August 2013, and the CIRFT was installed in the hot cell in September 2013 to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The fatigue responses of Zircaloy-4 (Zry-4) cladding and the role of pellet pellet and pellet clad interactions are critical to SNF vibration integrity, but such data are not available due to the unavailability of an effective testing system. While the deployment of the developed CIRFT test system in a hot cell will provide the opportunity to generate the data, the use of a surrogate rod has proven quite effective in identifying the underlying deformation mechanism of an SNF composite rod under an equivalent loading condition. This paper presents the experimental results of using surrogate rods under CIRFT reversible cyclic loading. Specifically, monotonic and cyclic bending tests were conducted on surrogate rods made of a Zry-4 tube and alumina pellet inserts, both with and without an epoxy bond.

  10. Experience with non-fuel-bearing components in LWR (light-water reactor) fuel systems

    SciTech Connect

    Bailey, W.J.; Berting, F.M.

    1990-12-01

    Many non-fuel-bearing components are so closely associated with the spent fuel assemblies that their integrity and behavior must be taken into consideration with the fuel assemblies, when handling spent fuel of planning waste management activities. Presented herein is some of the experience that has been gained over the past two decades from non-fuel-bearing components in light-water reactors (LWRs), both pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). Among the most important of these components are the control rod systems, the absorber and burnable poison rods, and the fuel assembly channels. 15 refs., 5 figs., 2 tabs.

  11. Multidimensional Fuel Performance Code: BISON

    2014-09-03

    BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficientlymore » solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phase field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.« less

  12. Multidimensional Fuel Performance Code: BISON

    SciTech Connect

    2014-09-03

    BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficiently solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phase field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.

  13. Evaluation of Cadmium Ratio and Foil Activation Measurements for a Beryllium-Reflected Assembly of U(93.15)O2 Fuel Rods (1.506-cm Triangular Pitch)

    SciTech Connect

    Marshall, Margaret A.

    2014-11-04

    A series of small, compact critical assembly (SCCA) experiments were completed from 1962 to 1965 at Oak Ridge National Laboratory’s Critical Experiments Facility (ORCEF) in support of the Medium-Power Reactor Experiments (MPRE) program. Initial experiments, performed in November and December of 1962, consisted of a core of un-moderated stainless-steel tubes, each containing 26 UOIdaho National Laboratory (INL), Idaho Falls, ID (United States) fuel pellets, surrounded by a graphite reflector. Measurements were performed to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. The graphite reflectors were then changed to beryllium reflectors. For the beryllium reflected assemblies, the fuel was in 1.506-cm-triangular and 7-tube clusters leading to two critical configurations. 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 1.506-cm-array critical configuration, have been evaluated and are described in this paper.

  14. Evaluation of Cadmium Ratio and Foil Activation Measurements for a Beryllium-Reflected Assembly of U(93.15)O2 Fuel Rods (1.506-cm Triangular Pitch)

    DOE PAGESBeta

    Marshall, Margaret A.

    2014-11-04

    A series of small, compact critical assembly (SCCA) experiments were completed from 1962 to 1965 at Oak Ridge National Laboratory’s Critical Experiments Facility (ORCEF) in support of the Medium-Power Reactor Experiments (MPRE) program. Initial experiments, performed in November and December of 1962, consisted of a core of un-moderated stainless-steel tubes, each containing 26 UOIdaho National Laboratory (INL), Idaho Falls, ID (United States) fuel pellets, surrounded by a graphite reflector. Measurements were performed to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. The graphite reflectors were then changed to beryllium reflectors. For the beryllium reflected assemblies, the fuel wasmore » in 1.506-cm-triangular and 7-tube clusters leading to two critical configurations. 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 1.506-cm-array critical configuration, have been evaluated and are described in this paper.« less

  15. Identification of failed fuel element

    DOEpatents

    Fryer, Richard M.; Matlock, Robert G.

    1976-06-22

    A passive fission product gas trap is provided in the upper portion of each fuel subassembly in a nuclear reactor. The gas trap consists of an inverted funnel of less diameter than the subassembly having a valve at the apex thereof. An actuating rod extends upwardly from the valve through the subassembly to a point where it can be contacted by the fuel handling mechanism for the reactor. Interrogation of the subassembly for the presence of fission products is accomplished by lowering the fuel handling machine onto the subassembly to press down on the actuating rod and open the valve.

  16. Prediction of number of breached rods following a LBLOCA of Candu plants using a BEPU approach

    SciTech Connect

    Bang, Y. S.; Kim, K.; Seul, K. W.; Woo, S. W.; Han, B. S.

    2012-07-01

    Radioactive doses following design basis accidents (DBA) have been important safety criteria of Candu nuclear power plant and they have been predicted in terms of the number of breached fuel rods. To support the licensing review on this concern, an analysis of LBLOCA has been conducted by using the BEPU method of KINS, KINS-REM. Number of Breached Rods (NBR) following a LBLOCA was predicted at 95 percentile probabilistic upper level in 95 percentile confidence level. Peak Cladding Temperatures (PCT) of the 84 bundles in the core pass 4 were calculated from the 124 MARS code runs in which the uncertainties of 10 major parameters including fuel thermal conductivity and break flow model were implemented. The fuel rod breaching criteria, PCT>1477 K, was used to determine the NBR 95/95. From the calculation, the predicted NBR 95/95 was 1591 rods and the calculated maximum NBR was lower than 2000 rods. Through the further improvements in feedback of the channel power behavior to thermalhydraulic calculation and in channel group modeling, NBR in more reliable level can be expected. (authors)

  17. 40 CFR Appendix C to Part 72 - Actual 1985 Yearly SO2 Emissions Calculation

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... PROGRAMS (CONTINUED) PERMITS REGULATION Pt. 72, App. C Appendix C to Part 72—Actual 1985 Yearly SO2... = (coal SO2 emissions) + (oil SO2 emissions) (in tons) If gas is the only fuel, gas emissions...

  18. Northrop Triga facility decommissioning plan versus actual results

    SciTech Connect

    Gardner, F.W.

    1986-01-01

    This paper compares the Triga facility decontamination and decommissioning plan to the actual results and discusses key areas where operational activities were impacted upon by the final US Nuclear Regulatory Commission (NRC)-approved decontamination and decommissioning plan. Total exposures for fuel transfer were a factor of 4 less than planned. The design of the Triga reactor components allowed the majority of the components to be unconditionally released.

  19. High temperature control rod assembly

    DOEpatents

    Vollman, Russell E.

    1991-01-01

    A high temperature nuclear control rod assembly comprises a plurality of substantially cylindrical segments flexibly joined together in succession by ball joints. The segments are made of a high temperature graphite or carbon-carbon composite. The segment includes a hollow cylindrical sleeve which has an opening for receiving neutron-absorbing material in the form of pellets or compacted rings. The sleeve has a threaded sleeve bore and outer threaded surface. A cylindrical support post has a threaded shaft at one end which is threadably engaged with the sleeve bore to rigidly couple the support post to the sleeve. The other end of the post is formed with a ball portion. A hollow cylindrical collar has an inner threaded surface engageable with the outer threaded surface of the sleeve to rigidly couple the collar to the sleeve. the collar also has a socket portion which cooperates with the ball portion to flexibly connect segments together to form a ball and socket-type joint. In another embodiment, the segment comprises a support member which has a threaded shaft portion and a ball surface portion. The threaded shaft portion is engageable with an inner threaded surface of a ring for rigidly coupling the support member to the ring. The ring in turn has an outer surface at one end which is threadably engageably with a hollow cylindrical sleeve. The other end of the sleeve is formed with a socket portion for engagement with a ball portion of the support member. In yet another embodiment, a secondary rod is slidably inserted in a hollow channel through the center of the segment to provide additional strength. A method for controlling a nuclear reactor utilizing the control rod assembly is also included.

  20. Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

    DOE PAGESBeta

    Bakosi, J.; Christon, M. A.; Lowrie, R. B.; Pritchett-Sheats, L. A.; Nourgaliev, R. R.

    2013-07-12

    The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies. In order to understand the fluid dynamics of GTRF and to build an archival database of turbulence statistics for various configurations, implicit large-eddy simulations of time-dependent single-phase turbulent flow have been performed in 3 × 3 and 5 × 5 rod bundles with a single grid spacer. To assess the computational mesh and resolution requirements, a method for quantitative assessment of unstructured meshes with no-slip walls is described. The calculations have been carriedmore » out using Hydra-TH, a thermal-hydraulics code developed at Los Alamos for the Consortium for Advanced Simulation of Light water reactors, a United States Department of Energy Innovation Hub. Hydra-TH uses a second-order implicit incremental projection method to solve the singlephase incompressible Navier-Stokes equations. The simulations explicitly resolve the large scale motions of the turbulent flow field using first principles and rely on a monotonicity-preserving numerical technique to represent the unresolved scales. Each series of simulations for the 3 × 3 and 5 × 5 rod-bundle geometries is an analysis of the flow field statistics combined with a mesh-refinement study and validation with available experimental data. Our primary focus is the time history and statistics of the forces loading the fuel rods. These hydrodynamic forces are believed to be the key player resulting in rod vibration and GTRF wear, one of the leading causes for leaking nuclear fuel which costs power utilities millions of dollars in preventive measures. As a result, we demonstrate that implicit large-eddy simulation of rod-bundle flows is a viable way to calculate the excitation forces for the GTRF problem.« less

  1. On-line monitoring of control rod integrity in BWRs using a mass spectrometer

    NASA Astrophysics Data System (ADS)

    Larsson, I.; Loner, H.; Ammon, K.; Sihver, L.; Ledergerber, G.

    2013-01-01

    Surveillance of fuel and control rod integrity in the core of a boiling water reactor is essential for maintaining a safe and reliable operation. Control rods of a boiling water reactor are mainly filled with boron carbide as a neutron absorber. Due to the irradiation of boron with neutrons, a continuous production of lithium and helium will occur inside a control rod. Most of the created helium will be retained in the boron carbide lattice; however a small part will escape into the void volume of the control blade. Therefore the integrity of control rods during operation can efficiently be followed by on-line measurements of helium concentration in the reactor off-gas system using a mass spectrometer. Since helium is a fill gas in fuel rods, the same method is a useful early warning system for primary fuel failures. In this paper, we introduce an on-line helium detector system which is installed at the nuclear power plant in Leibstadt. Furthermore the measuring experiences of control rod failure detection at the plant are presented. Different causes of increased helium levels in the off-gas system have been distinguished. There are spontaneous helium releases as well as helium releases caused by changed conditions in the reactor (power reduction, control rod movement, etc.). Helium peaks can also be characterized according to the released amount of helium, the peak shape and the duration of the release, which leads to different interpretations of the release mechanisms. In addition, the measured amount of released helium from a 50 days period (280 l) is also compared to the calculated amount of produced helium from the washed out boron during the same time period (190 l).

  2. Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

    SciTech Connect

    Bakosi, J.; Christon, M. A.; Lowrie, R. B.; Pritchett-Sheats, L. A.; Nourgaliev, R. R.

    2013-07-12

    The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies. In order to understand the fluid dynamics of GTRF and to build an archival database of turbulence statistics for various configurations, implicit large-eddy simulations of time-dependent single-phase turbulent flow have been performed in 3 × 3 and 5 × 5 rod bundles with a single grid spacer. To assess the computational mesh and resolution requirements, a method for quantitative assessment of unstructured meshes with no-slip walls is described. The calculations have been carried out using Hydra-TH, a thermal-hydraulics code developed at Los Alamos for the Consortium for Advanced Simulation of Light water reactors, a United States Department of Energy Innovation Hub. Hydra-TH uses a second-order implicit incremental projection method to solve the singlephase incompressible Navier-Stokes equations. The simulations explicitly resolve the large scale motions of the turbulent flow field using first principles and rely on a monotonicity-preserving numerical technique to represent the unresolved scales. Each series of simulations for the 3 × 3 and 5 × 5 rod-bundle geometries is an analysis of the flow field statistics combined with a mesh-refinement study and validation with available experimental data. Our primary focus is the time history and statistics of the forces loading the fuel rods. These hydrodynamic forces are believed to be the key player resulting in rod vibration and GTRF wear, one of the leading causes for leaking nuclear fuel which costs power utilities millions of dollars in preventive measures. As a result, we demonstrate that implicit large-eddy simulation of rod-bundle flows is a viable way to calculate the excitation forces for the GTRF problem.

  3. Self-assembly of rod-coil-rod ABA-type triblock copolymers.

    PubMed

    Chen, Ji-Zhong; Sun, Zhao-Yan; Zhang, Cheng-Xiang; An, Li-Jia; Tong, Zhen

    2008-02-21

    Self-assembled behavior of symmetric ABA rod-coil-rod triblock copolymer melts is studied by applying self-consistent-field lattice techniques in three-dimensional space. The phase diagram is constructed to understand the effects of the chain architecture on the self-assembled behavior. Four stable structures are observed for the ABA rod-coil-rod triblock, i.e., spherelike, lamellar, gyroidlike, and cylindrical structures. Different from AB rod-coil diblock and BAB coil-rod-coil triblock copolymers, the lamellar structure observed in ABA rod-coil-rod triblock copolymer melts is not stable for high volume fraction of the rod component (f(rod)=0.8), which is attributed to the intramolecular interactions between the two rod blocks of the polymer chain. When 0.3rod)<0.7, the intramolecular interactions between the two rigid blocks of the polymer chain are decrease, which results in the occurrence of some interesting metastable mixed structures. These structures have not been observed in polymers containing only one rod block, such as a lamellar-alt-lamellar structure. The results are expected to provide guidance for the design of microstructures in experiments. PMID:18298171

  4. 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.

  5. The actual goals of geoethics

    NASA Astrophysics Data System (ADS)

    Nemec, Vaclav

    2014-05-01

    The most actual goals of geoethics have been formulated as results of the International Conference on Geoethics (October 2013) held at the geoethics birth-place Pribram (Czech Republic): In the sphere of education and public enlightenment an appropriate needed minimum know how of Earth sciences should be intensively promoted together with cultivating ethical way of thinking and acting for the sustainable well-being of the society. The actual activities of the Intergovernmental Panel of Climate Changes are not sustainable with the existing knowledge of the Earth sciences (as presented in the results of the 33rd and 34th International Geological Congresses). This knowledge should be incorporated into any further work of the IPCC. In the sphere of legislation in a large international co-operation following steps are needed: - to re-formulate the term of a "false alarm" and its legal consequences, - to demand very consequently the needed evaluation of existing risks, - to solve problems of rights of individuals and minorities in cases of the optimum use of mineral resources and of the optimum protection of the local population against emergency dangers and disasters; common good (well-being) must be considered as the priority when solving ethical dilemmas. The precaution principle should be applied in any decision making process. Earth scientists presenting their expert opinions are not exempted from civil, administrative or even criminal liabilities. Details must be established by national law and jurisprudence. The well known case of the L'Aquila earthquake (2009) should serve as a serious warning because of the proven misuse of geoethics for protecting top Italian seismologists responsible and sentenced for their inadequate superficial behaviour causing lot of human victims. Another recent scandal with the Himalayan fossil fraud will be also documented. A support is needed for any effort to analyze and to disclose the problems of the deformation of the contemporary

  6. Automatic safety rod for reactors. [LMFBR

    DOEpatents

    Germer, J.H.

    1982-03-23

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  7. Nondestructive post-irradiation examination of Loop-1, S1 and B1 rods

    SciTech Connect

    Bratton, R.L.

    1997-05-01

    As a part of the Pacific Northwest National Laboratory`s Tritium Target Development Program, eleven tritium target rods were irradiated in the Advanced Test Reactor located at the Idaho National Engineering and Environmental Laboratory during 1991. Both nondestructive and destructive post-irradiation examination on all eleven rods was planned under the Tritium Target Development Program. Funding for the program was reduced in 1991 resulting in the early removal of the program experiments before reaching their irradiation goals. Post-irradiation examination was only performed on one of the irradiated rods at the Pacific Northwest National Laboratory before the program was terminated in 1992. On December 6, 1995, the Secretary of Energy announced the pursuit of the Commercial Light-Water Reactor option for producing tritium establishing the Tritium Target Qualification Program at the Pacific Northwest National Laboratory. This program decided to pursue nondestructive and destructive post-irradiation examination of the ten remaining rods from the previous program. The ten rods comprise three experiments. The Loop-1 experiment irradiated eight target rods in a loop configuration for 217 irradiation days. The other two rods were irradiated in two separate irradiation experiments, designated as S1 and B1 for 143 effective full-power days, but at different power levels. After the ten rods were transferred from the ATR Canal to the Hot Fuels Examination Facility, the following examinations were performed: (1) visual examination and photography; (2) neutron radiography; (3) axial gamma scanning; (4) contact profilometry measurement; (5) bow and length measurements; (6) rod puncture and plenum gas analysis/measurement of plenum gas quantity; (7) void volume determination; and (8) internal pressure determination. This report presents the data collected during these examinations.

  8. Fuel cell stack monitoring and system control

    DOEpatents

    Keskula, Donald H.; Doan, Tien M.; Clingerman, Bruce J.

    2004-02-17

    A control method for monitoring a fuel cell stack in a fuel cell system in which the actual voltage and actual current from the fuel cell stack are monitored. A preestablished relationship between voltage and current over the operating range of the fuel cell is established. A variance value between the actual measured voltage and the expected voltage magnitude for a given actual measured current is calculated and compared with a predetermined allowable variance. An output is generated if the calculated variance value exceeds the predetermined variance. The predetermined voltage-current for the fuel cell is symbolized as a polarization curve at given operating conditions of the fuel cell.

  9. Comparisons of COBRA-SFS calculations with data from simulated sections of unconsolidated and consolidated BWR spent fuel. Final report

    SciTech Connect

    Cuta, J.M.; Creer, J.M.

    1986-08-01

    The COBRA-SFS code was used to calculate temperatures in test sections modeling consolidated and unconsolidated spent fuel rods. The test sections were two-feet long, oriented horizontally, and designed so that the predominate direction of heat flow was in the radial direction. The data consisted of temperature measurements obtained on the electrically heated rods in the test assemblies, and on the boundary walls of the test sections. The COBRA-SFS calculations were in excellent agreement with the measured data from the consolidated test sections. The temperature calculated for the unconsolidated tests were somewhat higher than the measured values. Sensitivity studies on the models for the various modes of heat transfer available in the test sections show that for the consolidated geometry, the COBRA-SFS results are in agreement with the data, and well within experimental error. For the unconsolidated geometry, the COBRA-SFS results are conservative due to an idealized representation of the actual geometry.

  10. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    SciTech Connect

    DIAMOND,D.J.; ARONSON,A.; JO,J.; AVVAKUMOV,A.; MALOFEEV,V.; SIDOROV,V.; FERRARESI,P.; GOUIN,C.; ANIEL,S.; ROYER,M.E.

    1999-10-01

    This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

  11. Growth and Morphology of Rod Eutectics

    SciTech Connect

    Jing Teng; Shan Liu; R. Trivedi

    2008-03-17

    The formation of rod eutectic microstructure is investigated systematically in a succinonitrile-camphor alloy of eutectic composition by using the directional solidification technique. A new rod eutectic configuration is observed in which the rods form with elliptical cylindrical shape. Two different orientations of the ellipse are observed that differ by a 90{sup o} rotation such that the major and the minor axes are interchanged. Critical experiments in thin samples, where a single layer of rods forms, show that the spacing and orientation of the elliptic rods are governed by the growth rate and the sample thickness. In thicker samples, multi layers of rods form with circular cross-section and the scaling law between the spacing and velocity predicted by the Jackson and Hunt model is validated. A theoretical model is developed for a two-dimensional array of elliptical rods that are arranged in a hexagonal or a square array, and the results are shown to be consistent with the experimental observations. The model of elliptic rods is also shown to reduce to that for the circular rod eutectic when the lengths of the two axes are equal, and to the lamellar eutectic model when one of the axes is much larger than the other one.

  12. Temperature actuated automatic safety rod release

    DOEpatents

    Hutter, Ernest; Pardini, John A.; Walker, David E.

    1987-01-01

    A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

  13. Temperature actuated automatic safety rod release

    DOEpatents

    Hutter, E.; Pardini, J.A.; Walker, D.E.

    1984-03-13

    A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

  14. The directional sensitivity of retinal rods.

    PubMed Central

    Alpern, M; Ching, C C; Kitahara, K

    1983-01-01

    Rod field sensitivity, 10-S(r) (i.e. the reciprocal of the radiance of a background required for 10-fold elevation of rod threshold) was measured for monochromatic backgrounds traversing the pupil at various points (r) on three subjects. The wave-length dependency of the directional sensitivities of the three foveal cone mechanisms of the principal subject have been reported previously (Alpern & Kitahara, 1983). Rods, as cones, are less sensitive to obliquely incident, than to normally incident backgrounds. At the pupil margin (4 mm) the effect is between 0.368 and 0.976 log10 units smaller for rods. After correction for losses by corneal reflexion and by absorption in the lens, S(r) for rods is reasonably described by the parabolic equation used by Stiles (1937) to quantify the directional sensitivity of cones. The small effect for rods precludes a description as consistently precise as this equation provides for cones. The steepness of the parabolic curve best fitting the directional sensitivity data of the rods of the principal subject was independent of background wave number. For a second subject, whose rods are supposed to be smaller, it was directly proportional to the square of that wave number. The latter is the expectation if the directional sensitivity of this subject's rods were determined by principles outlined in the diffraction theory of Simon (1970). PMID:6644624

  15. Control Rod Pattern Planning of a BWR using Enhanced Nelder-Mead Method

    SciTech Connect

    Yoko Kobayashi; Eitaro Aiyoshi

    2004-07-01

    We propose a new optimization algorithm for the short-term planning of control rod patterns in an operating BWR. This algorithm is based on the enhanced Nelder-Mead simplex method in which convergence ability is improved for constrained problems in several ways. The main characteristic of this approach is it uses continuous values for the axial positions of control rods. Through calculations in an actual BWR plant, we showed that the new algorithm is effective for automation of short-term planning and reduction of the engineer's workload. (authors)

  16. Code System for Spent Fuel Heating Analysis.

    1999-05-24

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

  17. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, M.L.; Rosenstein, R.G.

    1998-10-13

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

  18. Mox fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-05-15

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

  19. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    1998-01-01

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  20. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-07-17

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  1. Connexin 36 and rod bipolar cell independent rod pathways drive retinal ganglion cells and optokinetic reflexes.

    PubMed

    Cowan, Cameron S; Abd-El-Barr, Muhammad; van der Heijden, Meike; Lo, Eric M; Paul, David; Bramblett, Debra E; Lem, Janis; Simons, David L; Wu, Samuel M

    2016-02-01

    Rod pathways are a parallel set of synaptic connections which enable night vision by relaying and processing rod photoreceptor light responses. We use dim light stimuli to isolate rod pathway contributions to downstream light responses then characterize these contributions in knockout mice lacking rod transducin-α (Trα), or certain pathway components associated with subsets of rod pathways. These comparisons reveal that rod pathway driven light sensitivity in retinal ganglion cells (RGCs) is entirely dependent on Trα, but partially independent of connexin 36 (Cx36) and rod bipolar cells. Pharmacological experiments show that rod pathway-driven and Cx36-independent RGC ON responses are also metabotropic glutamate receptor 6-dependent. To validate the RGC findings in awake, behaving animals we measured optokinetic reflexes (OKRs), which are sensitive to changes in ON pathways. Scotopic OKR contrast sensitivity was lost in Trα(-/-) mice, but indistinguishable from controls in Cx36(-/-) and rod bipolar cell knockout mice. Mesopic OKRs were also altered in mutant mice: Trα(-/-) mice had decreased spatial acuity, rod BC knockouts had decreased sensitivity, and Cx36(-/-) mice had increased sensitivity. These results provide compelling evidence against the complete Cx36 or rod BC dependence of night vision's ON component. Further, the findings suggest the parallel nature of rod pathways provides considerable redundancy to scotopic light sensitivity but distinct contributions to mesopic responses through complicated interactions with cone pathways. PMID:26718442

  2. Verification of the BISON fuel performance code

    SciTech Connect

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

    2014-09-01

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

  3. Fuel assembly self-excited vibration and test methodology

    SciTech Connect

    Lu, R.Y.; Broach, K. D.; McEvoy, J. J.

    2004-07-01

    PWR fuel assemblies normally experience low amplitude, random vibration under normal reactor flow conditions. This normal fuel assembly vibration has almost no impact on grid-rod fretting wear. However, some fuel assembly designs experience a high resonant fuel assembly vibration under normal axial flow conditions. This anomalous fuel assembly vibration is defined as fuel assembly self-excitation vibration (FASE), because the assembly vibrates resonantly without any external periodic excitation force. Fuel assembly self-excitation vibration can cause severe grid-rod fretting if the assembly operates at the flow rate, which causes high fuel assembly vibration. This paper will describe the characteristics of fuel assembly self-excitation vibration and the test methodology to identify the fuel assembly vibration. Several fuel assembly designs are compared under standard test conditions. The causes for the fuel assembly self-excitation vibration are analyzed and discussed. The test acceptance criteria are defined for newly developed PWR fuel assemblies. (authors)

  4. Fuels for sodium-cooled fast reactors: US perspective

    NASA Astrophysics Data System (ADS)

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.

    2007-09-01

    The US experience with mixed oxide, metal, and mixed carbide fuels is substantial, comprised of irradiation of over 50 000 MOX rods, over 130 000 metal rods, and 600 mixed carbide rods, in EBR-II and FFTF alone. All three types have been demonstrated capable of fuel utilization at or above 200 GWd/MTHM. To varying degrees, life-limiting phenomena for each type have been identified and investigated, and there are no disqualifying safety-related fuel behaviors. All three fuel types appear capable of meeting requirements of sodium-cooled fast reactor fuels, with reliability of mixed oxide and metal fuel well established. Improvements in irradiation performance of cladding and duct alloys have been a key development in moving these fuel designs toward higher-burnup potential. Selection of one fuel system over another will depend on circumstances particular to the application and on issues other than fuel performance, such as fabrication cost or overall system safety performance.

  5. PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

    SciTech Connect

    Loftus, M J; Hochreiter, L E; McGuire, M F; Valkovic, M M

    1983-10-01

    This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

  6. Computing Temperatures In Optically Pumped Laser Rods

    NASA Technical Reports Server (NTRS)

    Farrukh, Usamah O.

    1991-01-01

    Computer program presents new model solving temperature-distribution problem for laser rods of finite length and calculates both radial and axial components of temperature distributions in these rods. Contains several self-checking schemes to prevent over-writing of memory blocks and to provide simple tracing of information in case of trouble. Written in Microsoft FORTRAN 77.

  7. Tipping Time of a Quantum Rod

    ERIC Educational Resources Information Center

    Parrikar, Onkar

    2010-01-01

    The behaviour of a quantum rod, pivoted at its lower end on an impenetrable floor and restricted to moving in the vertical plane under the gravitational potential, is studied analytically under the approximation that the rod is initially localized to a "small-enough" neighbourhood around the point of classical unstable equilibrium. It is shown…

  8. 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.

  9. Vortex Noise from Rotating Cylindrical Rods

    NASA Technical Reports Server (NTRS)

    Stowell, E Z; Deming, A F

    1935-01-01

    A series of round rods of the some diameter were rotated individually about the mid-point of each rod. Vortices are shed from the rods when in motion, giving rise to the emission of sound. With the rotating system placed in the open air, the distribution of sound in space, the acoustical power output, and the spectral distribution have been studied. The frequency of emission of vortices from any point on the rod is given by the formula von Karman. From the spectrum estimates are made of the distribution of acoustical power along the rod, the amount of air concerned in sound production, the "equivalent size" of the vortices, and the acoustical energy content for each vortex.

  10. Power Histories for Fuel Codes

    SciTech Connect

    Gilbert, E. R.; Rausch, W. N.; Panisko, F. E.

    1982-01-01

    Computations of power history effects on the pre-loss-of-coolant accident (LOCA) conditions of generic pressurized water reactor (PWR) and boiling water reactor (BWR) fuel rods were performed at Pacific Northwest Laboratory using the U.S. Nuclear Regulatory Commission (NRC) code FRAPCON-2. Comparisons were made between cases where the fuel operated at a high ( 11 LOCA-limited") power throughout life (20,000 MWd/MTU) and those where the fuel was at a lower power for most of its burnup and ramped to the high power at 10,000 or 20,000 MWd/MTU burnup. The PWR rod was calculated to have more cladding creepdown during the lower power cases, which resulted in slightly lower centerline temperatures (as much as 100{degrees}C). This result was insensitive to the method used to increase the power during the ramps (i.e., by increasing the average rod power or by changing the peak-to-average (P/A} ratio of the axial power shape). The calculations also indicate that the highest fuel centerline temperatures were reached at startup. The BWR rod, however, demonstrated a substantial dependence on the power history. In this case, the constant high-power rod released considerably more fission gas than the lower power cases (21% versus 0.4%), which resulted in temperature differences of up to 350°C. The hiqhest temperature was reached at end-of-life (EOL) in the constant high-power case.

  11. Maintaining a Critical Spectra within Monteburns for a Gas-Cooled Reactor Array by Way of Control Rod Manipulation

    DOE PAGESBeta

    Adigun, Babatunde John; Fensin, Michael Lorne; Galloway, Jack D.; Trellue, Holly Renee

    2016-06-07

    Our burnup study examined the effect of a predicted critical control rod position on the nuclide predictability of several axial and radial locations within a 4×4 graphite moderated gas cooled reactor fuel cluster geometry. To achieve this, a control rod position estimator (CRPE) tool was developed within the framework of the linkage code Monteburns between the transport code MCNP and depletion code CINDER90, and four methodologies were proposed within the tool for maintaining criticality. Two of the proposed methods used an inverse multiplication approach - where the amount of fissile material in a set configuration is slowly altered until criticalitymore » is attained - in estimating the critical control rod position. Another method carried out several MCNP criticality calculations at different control rod positions, then used a linear fit to estimate the critical rod position. The final method used a second-order polynomial fit of several MCNP criticality calculations at different control rod positions to guess the critical rod position. The results showed that consistency in prediction of power densities as well as uranium and plutonium isotopics was mutual among methods within the CRPE tool that predicted critical position consistently well. Finall, while the CRPE tool is currently limited to manipulating a single control rod, future work could be geared toward implementing additional criticality search methodologies along with additional features.« less

  12. Hydraulic reinforcement of channel at lower tie-plate in BWR fuel bundle

    SciTech Connect

    Johansson, E.B.

    1989-12-26

    This patent describes an apparatus in a fuel bundle for confining fuel rods for the generation of steam in a steam water mixture passing interior of the fuel bundle. The fuel bundle includes: a lower tie-plate for supporting the fuel rods and permitting flow from the lower exterior portion of the fuel bundle into the interior portion of the fuel bundle; a plurality of fuel rods. The fuel rods supported on the lower tie-plate extending upwardly to and towards the upper portion of the fuel bundle for the generation of steam in a passing steam and water mixture interior of the fuel bundle; an upper tie-plate for maintaining the fuel rods in side-by-side relation and permitting a threaded connection between a plurality of the fuel rods with the threaded connection being at the upper and lower tie-plate. The upper tie-plate permitting escape of a steam water mixture from the top of the fuel bundle; a fuel bundle channel; and a labyrinth seal configured in the lower tie-plate.

  13. Improved Prediction of the Doppler Effect in TRISO Fuel

    SciTech Connect

    J. Ortensi; A.M. Ougouag

    2009-05-01

    The Doppler feedback mechanism is a major contributor to the passive safety of gas-cooled, graphite-moderated High Temperature Reactors that use fuel based on TRISO particles. It follows that the correct prediction of the magnitude and time-dependence of this feedback effect is essential to the conduct of safety analyses for these reactors. Since the effect is directly dependent on the actual temperature reached by the fuel during transients, the underlying phenomena of heat transfer and temperature rise must be correctly predicted. This paper presents an improved model for the TRISO particle and its thermal behavior during transients. The improved approach incorporates an explicit TRISO heat conduction model to better quantify the time dependence of the temperature in the various layers of the TRISO particle, including its fuel central zone. There follows a better treatment of the Doppler Effect within said fuel zone. The new model is based on a 1-D analytic solution for composite media using the Green’s function technique. The modeling improvement takes advantage of some of the physical behavior of TRISO fuel under irradiation and includes a distinctive look at the physics of the neutronic Doppler Effect. The new methodology has been implemented within the coupled R-Z nodal diffusion code CYNOD-THERMIX. The new model has been applied to the analysis of earthquakes (presented in a companion paper). In this paper, the model is applied to the control rod ejection event, as specified in the OECD PBMR-400 benchmark, but with temperature dependent thermal properties. The results obtained for this transient using the enhanced code are a considerable improvement over the predictions of the original code. The incorporation of the enhanced model shows that the Doppler Effect plays a more significant role than predicted by the original unenhanced model based on the THERMIX homogenized fuel region model. The new model shows that the overall energy generation during the rod

  14. Eulerian Formulation of Spatially Constrained Elastic Rods

    NASA Astrophysics Data System (ADS)

    Huynen, Alexandre

    Slender elastic rods are ubiquitous in nature and technology. For a vast majority of applications, the rod deflection is restricted by an external constraint and a significant part of the elastic body is in contact with a stiff constraining surface. The research work presented in this doctoral dissertation formulates a computational model for the solution of elastic rods constrained inside or around frictionless tube-like surfaces. The segmentation strategy adopted to cope with this complex class of problems consists in sequencing the global problem into, comparatively simpler, elementary problems either in continuous contact with the constraint or contact-free between their extremities. Within the conventional Lagrangian formulation of elastic rods, this approach is however associated with two major drawbacks. First, the boundary conditions specifying the locations of the rod centerline at both extremities of each elementary problem lead to the establishment of isoperimetric constraints, i.e., integral constraints on the unknown length of the rod. Second, the assessment of the unilateral contact condition requires, in principle, the comparison of two curves parametrized by distinct curvilinear coordinates, viz. the rod centerline and the constraint axis. Both conspire to burden the computations associated with the method. To streamline the solution along the elementary problems and rationalize the assessment of the unilateral contact condition, the rod governing equations are reformulated within the Eulerian framework of the constraint. The methodical exploration of both types of elementary problems leads to specific formulations of the rod governing equations that stress the profound connection between the mechanics of the rod and the geometry of the constraint surface. The proposed Eulerian reformulation, which restates the rod local equilibrium in terms of the curvilinear coordinate associated with the constraint axis, describes the rod deformed configuration

  15. Assessment of existing Sierra/Fuego capabilities related to grid-to-rod-fretting (GTRF).

    SciTech Connect

    Turner, Daniel Zack; Rodriguez, Salvador B.

    2011-06-01

    The following report presents an assessment of existing capabilities in Sierra/Fuego applied to modeling several aspects of grid-to-rod-fretting (GTRF) including: fluid dynamics, heat transfer, and fluid-structure interaction. We compare the results of a number of Fuego simulations with relevant sources in the literature to evaluate the accuracy, efficiency, and robustness of using Fuego to model the aforementioned aspects. Comparisons between flow domains that include the full fuel rod length vs. a subsection of the domain near the spacer show that tremendous efficiency gains can be obtained by truncating the domain without loss of accuracy. Thermal analysis reveals the extent to which heat transfer from the fuel rods to the coolant is improved by the swirling flow created by the mixing vanes. Lastly, coupled fluid-structure interaction analysis shows that the vibrational modes of the fuel rods filter out high frequency turbulent pressure fluctuations. In general, these results allude to interesting phenomena for which further investigation could be quite fruitful.

  16. CASL Virtual Reactor Predictive Simulation: Grid-to-Rod Fretting Wear

    SciTech Connect

    Roger, Lu Y.; Karoutas, Zeses; Sham, Sam

    2011-01-01

    Grid-to-Rod Fretting (GTRF) wear is currently one of the main causes of fuel rod leaking in pressurized water reactors. The Consortium for Advanced Simulation of Light Water Reactors (CASL) has identified GTRF as one of the Challenge Problems that drive the requirement for the development and application of a modeling and simulation computational environment for predictive simulation of light water reactors. This paper presents fretting wear simulation methodology currently employed by Westinghouse, a CASL industrial partner, to address GTRF. The required advancements in the computational and materials science modeling areas to develop a predictive simulation environment by CASL to address GTRF are outlined.

  17. Evaluation of differential shim rod worth measurements in the Oak Ridge Research Reactor

    SciTech Connect

    Bretscher, M.M.

    1987-01-01

    Reasonable agreement between calculated and measured differential shim rod worths in the Oak Ridge Research Reactor (ORR) has been achieved by taking into account the combined effects of negative reactivity contributions from changing fuel-moderator temperatures and of delayed photoneutrons. A method has been developed for extracting the asymptotic period from the shape of the initial portion of the measured time-dependent neutron flux profile following a positive reactivity insertion. In this region of the curve temperature-related reactivity feedback effects are negligibly small. Results obtained by applying this technique to differential shim rod worth measurements made in a wide variety of ORR cores are presented.

  18. Control Rod Malfunction at the NRAD Reactor

    SciTech Connect

    Thomas L. Maddock

    2010-05-01

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  19. DEVICE FOR CONTROLLING INSERTION OF ROD

    DOEpatents

    Beaty, B.J.

    1958-10-14

    A device for rapidly inserting a safety rod into a nuclear reactor upon a given signal or in the event of a power failure in order to prevent the possibility of extensive damage caused by a power excursion is described. A piston is slidably mounted within a vertical cylinder with provision for an electromagnetic latch at the top of the cylinder. This assembly, with a safety rod attached to the piston, is mounted over an access port to the core region of the reactor. The piston is normally latched at the top of the cylinder with the safety rod clear of the core area, however, when the latch is released, the piston and rod drop by their own weight to insert the rod. Vents along the side of the cylinder permit the escape of the air entrapped under the piston over the greater part of the distance, however, at the end of the fall the entrapped air is compressed thereby bringing the safety rod gently to rest, thus providing for a rapid automatic insertion of the rod with a minimum of structural shock.

  20. Granular materials interacting with thin flexible rods

    NASA Astrophysics Data System (ADS)

    Neto, Alfredo Gay; Campello, Eduardo M. B.

    2016-01-01

    In this work, we develop a computational model for the simulation of problems wherein granular materials interact with thin flexible rods. We treat granular materials as a collection of spherical particles following a discrete element method (DEM) approach, while flexible rods are described by a large deformation finite element (FEM) rod formulation. Grain-to-grain, grain-to-rod, and rod-to-rod contacts are fully permitted and resolved. A simple and efficient strategy is proposed for coupling the motion of the two types (discrete and continuum) of materials within an iterative time-stepping solution scheme. Implementation details are shown and discussed. Validity and applicability of the model are assessed by means of a few numerical examples. We believe that robust, efficiently coupled DEM-FEM schemes can be a useful tool to the simulation of problems wherein granular materials interact with thin flexible rods, such as (but not limited to) bombardment of grains on beam structures, flow of granular materials over surfaces covered by threads of hair in many biological processes, flow of grains through filters and strainers in various industrial segregation processes, and many others.

  1. Gelation and mechanical response of patchy rods.

    PubMed

    Kazem, Navid; Majidi, Carmel; Maloney, Craig E

    2015-10-28

    We perform Brownian dynamics simulations to study the gelation of suspensions of attractive, rod-like particles. We show that in detail the rod-rod surface interactions can dramatically affect the dynamics of gelation and the structure and mechanics of the networks that form. If the attraction between the rods is perfectly smooth along their length, they will collapse into compact bundles. If the attraction is sufficiently corrugated or patchy, over time, a rigid space-spanning network will form. We study the structure and mechanical properties of the networks that form as a function of the fraction of the surface, f, that is allowed to bind. Surprisingly, the structural and mechanical properties are non-monotonic in f. At low f, there are not a sufficient number of cross-linking sites to form networks. At high f, rods bundle and form disconnected clusters. At intermediate f, robust networks form. The elastic modulus and yield stress are both non-monotonic in the surface coverage. The stiffest and strongest networks show an essentially homogeneous deformation under strain with rods re-orienting along the extensional axis. Weaker, more clumpy networks at high f re-orient relatively little with strong non-affine deformation. These results suggest design strategies for tailoring surface interactions between rods to yield rigid networks with optimal mechanical properties. PMID:26381995

  2. More on the penetration of yawed rods

    NASA Astrophysics Data System (ADS)

    Rosenberg, Z.; Dekel, E.; Ashuach, Y.

    2006-08-01

    One of the most complex processes, in the field of terminal ballistics, is that of yawed impact of long rods. In spite of many experimental observations, and some analytical modeling, a clear picture of this issue is still lacking. In order to gain some insight into the operating mechanisms, we developed a simple engineering model which considers the yawed rod as a series of small disks. We then define the effective length and diameter of the rod by considering those disks which are going to hit the initial crater which is opened by the impact. We also performed a series of 3D numerical simulations with various L/D tungsten alloy rods impacting a steel target, at yaws in the full range of 0-90^circ. We analyzed the results of these simulations in terms of the normalized penetration (P/D), where D is the rod diameter, and looked for systematic trends in the results for the various rods. The agreement between our model predictions and both experimental data and simulation results is quite good. Based on this agreement we can highlight some new features of the penetration process of yawed rods.

  3. Dynamic response of RC beams strengthened with near surface mounted Carbon-FRP rods subjected to damage

    NASA Astrophysics Data System (ADS)

    Capozucca, R.; Blasi, M. G.; Corina, V.

    2015-07-01

    Near surface mounted (NSM) technique with fiber reinforced polymer (FRP) is becoming a common method in the strengthening of concrete beams. The availability of NSM FRP technique depends on many factors linked to materials and geometry - dimensions of the rods used, type of FRP material employed, rods’ surface configuration, groove size - and to adhesion between concrete and FRP rods. In this paper detection of damage is investigated measuring the natural frequency values of beam in the case of free-free ends. Damage was due both to reduction of adhesion between concrete and carbon-FRP rectangular and circular rods and cracking of concrete under static bending tests on beams. Comparison between experimental and theoretical frequency values evaluating frequency changes due to damage permits to monitor actual behaviour of RC beams strengthened by NSM CFRP rods.

  4. Phototransduction in mouse rods and cones

    PubMed Central

    Fu, Yingbin; Yau, King-Wai

    2010-01-01

    Phototransduction is the process by which light triggers an electrical signal in a photoreceptor cell. Image-forming vision in vertebrates is mediated by two types of photoreceptors: the rods and the cones. In this review, we provide a summary of the success in which the mouse has served as a vertebrate model for studying rod phototransduction, with respect to both the activation and termination steps. Cones are still not as well-understood as rods partly because it is difficult to work with mouse cones due to their scarcity and fragility. The situation may change, however. PMID:17226052

  5. Control rod drive hydraulic system

    DOEpatents

    Ose, Richard A.

    1992-01-01

    A hydraulic system for a control rod drive (CRD) includes a variable output-pressure CR pump operable in a charging mode for providing pressurized fluid at a charging pressure, and in a normal mode for providing the pressurized fluid at a purge pressure, less than the charging pressure. Charging and purge lines are disposed in parallel flow between the CRD pump and the CRD. A hydraulic control unit is disposed in flow communication in the charging line and includes a scram accumulator. An isolation valve is provided in the charging line between the CRD pump and the scram accumulator. A controller is operatively connected to the CRD pump and the isolation valve and is effective for opening the isolation valve and operating the CRD pump in a charging mode for charging the scram accumulator, and closing the isolation valve and operating the CRD pump in a normal mode for providing to the CRD through the purge line the pressurized fluid at a purge pressure lower than the charging pressure.

  6. Description of Axial Detail for ROK Fuel

    SciTech Connect

    Trellue, Holly R; Galloway, Jack D

    2012-04-20

    For the purpose of NDA simulations of the ROK fuel assemblies, we have developed an axial burnup distribution to represent the pins themselves based on gamma scans of rods in the G23 assembly. For the purpose of modeling the G23 assembly (both at ORNL and LANL), the pin-by-pin burnup map as simulated by ROK is being assumed to represent the radial burnup distribution. However, both DA and NDA results indicate that this simulated estimate is not 100% correct. In particular, the burnup obtained from the axial gamma scan of 7 pins does not represent exactly the same 'average' pin burnup as the ROK simulation. Correction for this discrepancy is a goal of the well-characterized assembly task but will take time. For now, I have come up with a correlation for 26 axial points of the burnup as obtained by gamma scans of 7 different rods (C13, G01, G02, J11, K10, L02, and M04, neglecting K02 at this time) to the average burnup given by the simulation for each of the rods individually. The resulting fraction in each axial zone is then averaged for the 7 different rods so that it can represent every fuel pin in the assembly. The burnup in each of the 26 axial zones of rods in all ROK assemblies will then be directly adjusted using this fraction, which is given in Table 1. Note that the gamma scan data given by ROK for assembly G23 included a length of {approx}3686 mm, so the first 12 mm and the last 14 mm were ignored to give an actual rod length of {approx}366 cm. To represent assembly F02 in which no pin-by-pin burnup distribution is given by ROK, we must model it using infinitely-reflected geometry but can look at the effects of measuring in different axial zones by using intermediate burnup files (i.e. smaller burnups than 28 GWd/MTU) and determining which axial zone(s) each burnup represents. Details for assembly F02 are then given in Tables 2 and 3, which is given in Table 1 and has 44 total axial zones to represent the top meter in explicit detail in addition to the

  7. Moral Reasoning in Hypothetical and Actual Situations.

    ERIC Educational Resources Information Center

    Sumprer, Gerard F.; Butter, Eliot J.

    1978-01-01

    Results of this investigation suggest that moral reasoning of college students, when assessed using the DIT format, is the same whether the dilemmas involve hypothetical or actual situations. Subjects, when presented with hypothetical situations, become deeply immersed in them and respond as if they were actual participants. (Author/BEF)

  8. Factors Related to Self-Actualization.

    ERIC Educational Resources Information Center

    Hogan, H. Wayne; McWilliams, Jettie M.

    1978-01-01

    Provides data to further support the notions that females score higher in self-actualization measures and that self-actualization scores correlate inversely to the degree of undesirability individuals assign to their heights and weights. Finds that, contrary to predictions, greater androgyny was related to lower, not higher, self-actualization…

  9. Pulse-actuated fuel-injection spark plug

    DOEpatents

    Murray, Ian; Tatro, Clement A.

    1978-01-01

    A replacement spark plug for reciprocating internal combustion engines that functions as a fuel injector and as a spark plug to provide a "stratified-charge" effect. The conventional carburetor is retained to supply the main fuel-air mixture which may be very lean because of the stratified charge. The replacement plug includes a cylindrical piezoelectric ceramic which contracts to act as a pump whenever an ignition pulse is applied to a central rod through the ceramic. The rod is hollow at its upper end for receiving fuel, it is tapered along its lower length to act as a pump, and it is flattened at its lower end to act as a valve for fuel injection from the pump into the cylinder. The rod also acts as the center electrode of the plug, with the spark jumping from the plug base to the lower end of the rod to thereby provide spark ignition that has inherent proper timing with the fuel injection.

  10. Power Delivery from an Actual Thermoelectric Generation System

    NASA Astrophysics Data System (ADS)

    Kaibe, Hiromasa; Kajihara, Takeshi; Nagano, Kouji; Makino, Kazuya; Hachiuma, Hirokuni; Natsuume, Daisuke

    2014-06-01

    Similar to photovoltaic (PV) and fuel cells, thermoelectric generators (TEGs) supply direct-current (DC) power, essentially requiring DC/alternating current (AC) conversion for delivery as electricity into the grid network. Use of PVs is already well established through power conditioning systems (PCSs) that enable DC/AC conversion with maximum-power-point tracking, which enables commercial use by customers. From the economic, legal, and regulatory perspectives, a commercial PCS for PVs should also be available for TEGs, preferably as is or with just simple adjustment. Herein, we report use of a PV PCS with an actual TEG. The results are analyzed, and proper application for TEGs is proposed.

  11. Irradiation Test of Advanced PWR Fuel in Fuel Test Loop at HANARO

    SciTech Connect

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo; Park, Su Ki; Seo, Chul Gyo

    2007-07-01

    A new fuel test loop has been constructed in the research reactor HANARO at KAERI. The main objective of the FTL (Fuel Test Loop) is an irradiation test of a newly developed LWR fuel under PWR or Candu simulated conditions. The first test rod will be loaded within 2007 and its irradiation test will be continued until a rod average their of 62 MWd/kgU. A total of five test rods can be loaded into the IPS (In-Pile Section) and fuel centerline temperature, rod internal pressure and fuel stack elongation can be measured by an on-line real time system. A newly developed advanced PWR fuel which consists of a HANA{sup TM} alloy cladding and a large grain UO{sub 2} pellet was selected as the first test fuel in the FTL. The fuel cladding, the HANA{sup TM} alloy, is an Nb containing Zirconium alloy that has shown better corrosion and creep resistance properties than the current Zircaloy-4 cladding. A total of six types of HANA{sup TM} alloy were developed and two or three of these candidate alloys will be used as test rod cladding, which have shown a superior performance to the others. A large-grain UO{sub 2} pellet has a 14{approx}16 micron 2D diameter grain size for a reduction of a fission gas release at a high burnup. In this paper, characteristics of the FTL and IPS are introduced and the expected operation and irradiation conditions are summarized for the test periods. Also the preliminary fuel performance analysis results, such as the cladding oxide thickness, fission gas release and rod internal pressure, are evaluated from the test rod safety analysis aspects. (authors)

  12. Validation of the BISON 3D Fuel Performance Code: Temperature Comparisons for Concentrically and Eccentrically Located Fuel Pellets

    SciTech Connect

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

    2013-03-01

    BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. The code is applicable to both steady and transient fuel behaviour and is used to analyse either 2D axisymmetric or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods. Halden IFA experiments constitute a large percentage of the current BISON validation base. The validation emphasis here is centreline temperatures at the beginning of fuel life, with comparisons made to seven rods from the IFA-431 and 432 assemblies. The principal focus is IFA-431 Rod 4, which included concentric and eccentrically located fuel pellets. This experiment provides an opportunity to explore 3D thermomechanical behaviour and assess the 3D simulation capabilities of BISON. Analysis results agree with experimental results showing lower fuel centreline temperatures for eccentric fuel with the peak temperature shifted from the centreline. The comparison confirms with modern 3D analysis tools that the measured temperature difference between concentric and eccentric pellets is not an artefact and provides a quantitative explanation for the difference.

  13. Decontamination of control rod housing from Palisades Nuclear Power Station.

    SciTech Connect

    Kaminski, M.D.; Nunez, L.; Purohit, A.

    1999-05-03

    Argonne National Laboratory has developed a novel decontamination solvent for removing oxide scales formed on ferrous metals typical of nuclear reactor piping. The decontamination process is based on the properties of the diphosphonic acids (specifically 1-hydroxyethane-1,1-diphosphonic acid or HEDPA) coupled with strong reducing-agents (e.g., sodium formaldehyde sulfoxylate, SFS, and hydroxylamine nitrate, HAN). To study this solvent further, ANL has solicited actual stainless steel piping material that has been recently removed from an operating nuclear reactor. On March 3, 1999 ANL received segments of control rod housing from Consumers Energy's Palisades Nuclear Plant (Covert, MI) containing radioactive contamination from both neutron activation and surface scale deposits. Palisades Power plant is a PWR type nuclear generating plant. A total of eight segments were received. These segments were from control rod housing that was in service for about 6.5 years. Of the eight pieces that were received two were chosen for our experimentation--small pieces labeled Piece A and Piece B. The wetted surfaces (with the reactor's pressurized water coolant/moderator) of the pieces were covered with as a scale that is best characterized visually as a smooth, shiny, adherent, and black/brown in color type oxide covering. This tenacious oxide could not be scratched or removed except by aggressive mechanical means (e.g., filing, cutting).

  14. High current pulse testing for ground rod integrity

    NASA Technical Reports Server (NTRS)

    Walko, Lawrence C.

    1991-01-01

    A test technique was developed to assess various grounding system concepts used for mobile facilities. The test technique involves applying a high current pulse to the grounding system with the proper waveshape and magnitude to simulate a lightning return stroke. Of concern were the step voltages present along the ground near the point of lightning strike. Step voltage is equated to how fast the current pulse is dissipated by the grounding system. The applied current pulse was produced by a high current capacitor bank with a total energy content of 80 kilojoules. A series of pulse tests were performed on two types of mobile facility grounding systems. One system consisted of an array of four 10 foot copper clad steel ground rods connected by 1/0 gauge wire. The other system was an array of 10 inch long tapered ground rods, strung on stainless steel cable. The focus here is on the pulse test technique used and its relevance to actual lightning strike conditions.

  15. Control rod for a nuclear reactor

    DOEpatents

    Roman, Walter G.; Sutton, Jr., Harry G.

    1979-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod.

  16. Impact of AD995 alumina rods

    SciTech Connect

    Chhabildas, L.C.; Furnish, M.D.; Reinhart, W.D.; Grady, D.E.

    1997-10-01

    Gas guns and velocity interferometric techniques have been used to determine the loading behavior of an AD995 alumina rod 19 mm in diameter by 75 mm and 150 mm long, respectively. Graded-density materials were used to impact both bare and sleeved alumina rods while the velocity interferometer was used to monitor the axial-velocity of the free end of the rods. Results of these experiments demonstrate that (1) a time-dependent stress pulse generated during impact allows an efficient transition from the initial uniaxial strain loading to a uniaxial stress state as the stress pulse propagates through the rod, and (2) the intermediate loading rates obtained in this configuration lie between split Hopkinson bar and shock-loading techniques.

  17. Catalytic combustion of actual low and medium heating value gases

    NASA Technical Reports Server (NTRS)

    Bulzan, D. L.

    1982-01-01

    Catalytic combustion of both low and medium heating value gases using actual coal derived gases obtained from operating gasifiers was demonstrated. A fixed bed gasifier with a complete product gas cleanup system was operated in an air blown mode to produce low heating value gas. A fluidized bed gasifier with a water quench product gas cleanup system was operated in both an air enriched and an oxygen blown mode to produce low and medium, heating value gas. Noble metal catalytic reactors were evaluated in 12 cm flow diameter test rigs on both low and medium heating value gases. Combustion efficiencies greater than 99.5% were obtained with all coal derived gaseous fuels. The NOx emissions ranged from 0.2 to 4 g NO2 kg fuel.

  18. Examination of the first irradiated LOFT fuel module

    SciTech Connect

    Cook, J.A.; Olsen, C.S.

    1981-01-01

    The first Loss-of-Fluid Test (LOFT) center fuel module was nondestructively examined in order to assess any changes after power range testing and three large-break loss-of-coolant experiments (LOCEs). The examination consisted of evaluation of LOCE measurement data; measurement of withdrawal forces during removal of the module from the reactor; poolside examination of the exposed fuel module surfaces, using an underwater periscope, 35-mm camera, and closed circuit television; and poolside measurements of the rod-to-rod spacing, using a Sulo probe. The performance of the equipment is assessed from the results of the examination. Color standards are required for underwater color photography, and fuel rod deflection must be considered in evaluting rod-to-rod spaces.

  19. Need and procedure for calibration of fuel rod simulators. [PWR

    SciTech Connect

    Dabbs, R.D.; Ott, L.J.

    1980-01-01

    An experimental thermocouple calibration procedure and four-part calibration program, ORTCAL (ORNL Thermocouple Calibration), have been developed to supply FRS performance information to the inverse heat conduction model and program ORINC. Case studies have shown that failure to fully classify FRSs with regard to component physical properties, gaps, etc., can result in severe errors during inverse calculations of the driving potential at the surface of the FRS (..delta..T), the spatial and temporal history of the heat flow within the FRS, and the surface heat flux.

  20. Summary report of fuel rod displacement sensor for LOFT

    SciTech Connect

    Billeter, T.R.

    1980-01-01

    Qualification tests conducted for a period of 700 hours of each of three displacement measuring (LVDT) sensors confirmed applicability of the design for use in the Loss-of-Fluid-Test (LOFT) reactor. Operationally, the sensor satisfies all specified requirements for LOFT. Even for imposed temperature transients at rates up to 100/sup 0/F/s, the indicated displacement remained within the allowed maximum error band of +-10% of reading. The 0.625-inch O.D. by 5.5-inch long sensor exhibited a linearly related signal output variation for displacement variations of up to 1-inch range. Long term operation at temperatures of 100/sup 0/F to 800/sup 0/F caused no perceptible permanent change of operating characteristics.

  1. Measurements of control rod efficiency in RBMK critical assembly upon dropping of the rods

    SciTech Connect

    Zhitarev, V. E. Kachanov, V. M.; Sergevnin, A. Yu.; Lebedev, G. V.

    2014-12-15

    The efficiency of control rods in the RBMK critical assembly was measured in the case where one manual-control rod (MCR) is dropped from a steady critical state, and several other MCRs were additionally dropped after 44 s. The measured number of neutrons in the assembly during and after dropping of the rods was used to calculate the efficiency values of the rods by solution of the system of point kinetics equations. A series of methods of the initial data treatment for determination of the desired values of reactivity without the calculated corrections were used.

  2. Epigenomic landscapes of retinal rods and cones

    PubMed Central

    Mo, Alisa; Luo, Chongyuan; Davis, Fred P; Mukamel, Eran A; Henry, Gilbert L; Nery, Joseph R; Urich, Mark A; Picard, Serge; Lister, Ryan; Eddy, Sean R; Beer, Michael A; Ecker, Joseph R; Nathans, Jeremy

    2016-01-01

    Rod and cone photoreceptors are highly similar in many respects but they have important functional and molecular differences. Here, we investigate genome-wide patterns of DNA methylation and chromatin accessibility in mouse rods and cones and correlate differences in these features with gene expression, histone marks, transcription factor binding, and DNA sequence motifs. Loss of NR2E3 in rods shifts their epigenomes to a more cone-like state. The data further reveal wide differences in DNA methylation between retinal photoreceptors and brain neurons. Surprisingly, we also find a substantial fraction of DNA hypo-methylated regions in adult rods that are not in active chromatin. Many of these regions exhibit hallmarks of regulatory regions that were active earlier in neuronal development, suggesting that these regions could remain undermethylated due to the highly compact chromatin in mature rods. This work defines the epigenomic landscapes of rods and cones, revealing features relevant to photoreceptor development and function. DOI: http://dx.doi.org/10.7554/eLife.11613.001 PMID:26949250

  3. Close packing of rods on spherical surfaces

    NASA Astrophysics Data System (ADS)

    Smallenburg, Frank; Löwen, Hartmut

    2016-04-01

    We study the optimal packing of short, hard spherocylinders confined to lie tangential to a spherical surface, using simulated annealing and molecular dynamics simulations. For clusters of up to twelve particles, we map out the changes in the geometry of the closest-packed configuration as a function of the aspect ratio L/D, where L is the cylinder length and D the diameter of the rods. We find a rich variety of cluster structures. For larger clusters, we find that the best-packed configurations up to around 100 particles are highly dependent on the exact number of particles and aspect ratio. For even larger clusters, we find largely disordered clusters for very short rods (L/D = 0.25), while slightly longer rods (L/D = 0.5 or 1) prefer a global baseball-like geometry of smectic-like domains, similar to the behavior of large-scale nematic shells. Intriguingly, we observe that when compared to their optimal flat-plane packing, short rods adapt to the spherical geometry more efficiently than both spheres and longer rods. Our results provide predictions for experimentally realizable systems of colloidal rods trapped at the interface of emulsion droplets.

  4. Close packing of rods on spherical surfaces.

    PubMed

    Smallenburg, Frank; Löwen, Hartmut

    2016-04-28

    We study the optimal packing of short, hard spherocylinders confined to lie tangential to a spherical surface, using simulated annealing and molecular dynamics simulations. For clusters of up to twelve particles, we map out the changes in the geometry of the closest-packed configuration as a function of the aspect ratio L/D, where L is the cylinder length and D the diameter of the rods. We find a rich variety of cluster structures. For larger clusters, we find that the best-packed configurations up to around 100 particles are highly dependent on the exact number of particles and aspect ratio. For even larger clusters, we find largely disordered clusters for very short rods (L/D = 0.25), while slightly longer rods (L/D = 0.5 or 1) prefer a global baseball-like geometry of smectic-like domains, similar to the behavior of large-scale nematic shells. Intriguingly, we observe that when compared to their optimal flat-plane packing, short rods adapt to the spherical geometry more efficiently than both spheres and longer rods. Our results provide predictions for experimentally realizable systems of colloidal rods trapped at the interface of emulsion droplets. PMID:27131565

  5. High-throughput rod-induced electrospinning

    NASA Astrophysics Data System (ADS)

    Wu, Dezhi; Xiao, Zhiming; Teh, Kwok Siong; Han, Zhibin; Luo, Guoxi; Shi, Chuan; Sun, Daoheng; Zhao, Jinbao; Lin, Liwei

    2016-09-01

    A high throughput electrospinning process, directly from flat polymer solution surfaces induced by a moving insulating rod, has been proposed and demonstrated. Different rods made of either phenolic resin or paper with a diameter of 1–3 cm and a resistance of about 100–500 MΩ, has been successfully utilized in the process. The rod is placed approximately 10 mm above the flat polymer solution surface with a moving speed of 0.005–0.4 m s‑1 this causes the solution to generate multiple liquid jets under an applied voltage of 15–60 kV for the tip-less electrospinning process. The local electric field induced by the rod can boost electrohydrodynamic instability in order to generate Taylor cones and liquid jets. Experimentally, it is found that a large rod diameter and a small solution-to-rod distance can enhance the local electrical field to reduce the magnitude of the applied voltage. In the prototype setup with poly (ethylene oxide) polymer solution, an area of 5 cm  ×  10 cm and under an applied voltage of 60 kV, the maximum throughput of nanofibers is recorded to be approximately144 g m‑2 h‑1.

  6. 49 CFR 230.93 - Pistons and piston rods.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Pistons and piston rods. 230.93 Section 230.93... Tenders Driving Gear § 230.93 Pistons and piston rods. (a) Maintenance and testing. Pistons and piston rods shall be maintained in safe and suitable condition for service. Piston rods shall be inspected...

  7. Mechanical performance of fiberglass sucker-rod strings

    SciTech Connect

    Tripp, H.A.

    1988-08-01

    The natural frequencies of fiberglass sucker-rod strings can be calculated by treating the rod strings as modified spring/mass vibration systems. The ratio of the pumping-unit operating speed to the rod-string natural frequency can then be used as a basis for understanding fiberglass-rod performance and for predicting downhole pump stroke lengths.

  8. Reassessment of the basis for NRC fuel damage criteria for reactivity transients

    SciTech Connect

    McCardell, R.K.

    1994-10-01

    The present basis for NRC Fuel Damage Criteria was obtained from experiments performed in the Special Power Excursion Reactor Test (SPERT) IV Reactor Capsule Driver Core (CDC) at the Idaho National Engineering Laboratory (INEL) between 1967 and 1970. Most of the CDC test fuel rods were previously unirradiated and the failure threshold for these unirradiated fuel rods was measured to be about 200 calories per gram of UO{sub 2} radially averaged fuel enthalpy at the axial peak.

  9. Feasibility study of the University of Utah TRIGA reactor power upgrade in respect to control rod system

    NASA Astrophysics Data System (ADS)

    Cutic, Avdo

    The objectives of this thesis are twofold: to determine the highest achievable power levels of the current University of Utah TRIG Reactor (UUTR) core configuration with the existing three control rods, and to design the core for higher reactor power by optimizing the control rod worth. For the current core configuration, the maximum reactor power, eigenvalue keff, shutdown margin, and excess reactivity have been measured and calculated. These calculated estimates resulted from thermal power calibrations, and the control rod worth measurements at various power levels. The results were then used as a benchmark to verify the MCNP5 core simulations for the current core and then to design a core for higher reactor power. This study showed that the maximum achievable power with the current core configuration and control rod system is 150kW, which is 50kW higher than the licensed power of the UUTR. The maximum achievable UUTR core power with the existing fuel is determined by optimizing the core configuration and control rod worth, showing that a power upgrade of 500 kW is achievable. However, it requires a new control rod system consisting of a total of four control rods. The cost of such an upgrade is $115,000.

  10. Nuclear reactor remote disconnect control rod coupling indicator

    DOEpatents

    Vuckovich, Michael

    1977-01-01

    A coupling indicator for use with nuclear reactor control rod assemblies which have remotely disengageable couplings between the control rod and the control rod drive shaft. The coupling indicator indicates whether the control rod and the control rod drive shaft are engaged or disengaged. A resistive network, utilizing magnetic reed switches, senses the position of the control rod drive mechanism lead screw and the control rod position indicating tube, and the relative position of these two elements with respect to each other is compared to determine whether the coupling is engaged or disengaged.

  11. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    NASA Astrophysics Data System (ADS)

    Lind, Terttaliisa; Csordás, Anna Pintér; Nagy, Imre; Stuckert, Juri

    2010-02-01

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T ˜ 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T ˜ 1650 K, followed by a relatively

  12. Rod-bundle transient-film boiling of high-pressure water in the liquid-deficient regime. [PWR

    SciTech Connect

    Morris, D.G.; Mullins, C.B.; Yoder, G.L.

    1982-01-01

    Results are reported from a recent experiment investigating dispersed flow film boiling of high pressure water in upflow through a rod bundle. The data, obtained under mildly transient conditions, are used to assess correlations currently used to predict heat transfer in these circumstances. In light of the scarcity of similar data, the data should prove useful in the development and assessment of new heat transfer models. The experiment was conducted at the Oak Ridge National Laboratory in the Thermal-Hydraulic Test Facility, a highly instrumented, non-nuclear, pressurized-water loop containing 64, 3.66-m (12-ft) long rods (of which 60 are electrically heated). The rods are arranged in a square array typical of 17 x 17 fuel rod assemblies in late generation PWRs. Data were collected over typical reactor blowdown parameter ranges.

  13. Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

    NASA Astrophysics Data System (ADS)

    Shoushtari, M. K.; Kakavand, T.; Sadat Kiai, S. M.; Ghaforian, H.

    2010-03-01

    The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity ( ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

  14. Protected Nuclear Fuel Element

    DOEpatents

    Kittel, J. H.; Schumar, J. F.

    1962-12-01

    A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

  15. Fuels and materials for LMFBR core components

    SciTech Connect

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

    1984-04-01

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

  16. TRIGA spent-fuel storage criticality analysis

    SciTech Connect

    Ravnik, M.; Glumac, B.

    1996-06-01

    A criticality safety analysis of a pool-type storage for spent TRIGA Mark II reactor fuel is presented. Two independent computer codes are applied: the MCNP Monte Carlo code and the WIMS lattice cell code. Two types of fuel elements are considered: standard fuel elements with 12 wt% uranium concentration and FLIP fuel elements. A parametric study of spent-fuel storage lattice pitch, fuel element burnup, and water density is presented. Normal conditions and postulated accident conditions are analyzed. A strong dependence of the multiplication factor on the distance between the fuel elements and on the effective water density is observed. A multiplication factor <1 may be expected for an infinite array of fuel rods at center-to-center distances >6.5 cm, regardless of the fuel element type and burnup. At shorter distances, the subcriticality can be ensured only by adding absorbers to the array of fuel rods even if the fuel rods were burned to {approximately}20% burnup. The results of both codes agree well for normal conditions. The results show that WIMS may be used as a complement to the Monte Carlo code in some parts of the criticality analysis.

  17. Rod Control Assemblies Wear Mechanisms

    SciTech Connect

    Kaczorowski, Damien; Georges, Jean-Mary; Bec, Sandrine; Vannes, Andre-Bernard; Tonck, Andre; Vernot, Jean-Philippe

    2002-07-01

    In nuclear power plants, slender tubular components are subjected to vibrations in a PHTW environment. As a result, the two contacting surfaces, tubes and their guides undergo impact at low contact pressures. The components are usually made of stainless steel and it was found that the influence of the PHTW, combined with other actions (such as corrosion, erosion, squeeze film effect, third body effect and cavitation) leads to a particular wear of the material. Therefore, this paper aims to show that the colloidal oxides, formed on the steel surfaces in PHTW, play a principal role in the wear of the surfaces. Actually, due to the specific kinematic conditions of the contact, the flow of compacted oxides abrades the surfaces. (authors)

  18. A GAMMA RAY SCANNING APPROACH TO QUANTIFY SPENT FUEL CASK RADIONUCLIDE CONTENTS

    SciTech Connect

    Branney, S.

    2011-07-01

    The International Atomic Energy Agency (IAEA) has outlined a need to develop methods of allowing re-verification of LWR spent fuel stored in dry storage casks without the need of a reference baseline measurement. Some scanning methods have been developed, but improvements can be made to readily provide required data for spent fuel cask verification. The scanning process should be conditioned to both confirm the contents and detect any changes due to container/contents degradation or unauthorized removal or tampering. Savannah River National Laboratory and The University of Tennessee are exploring a new method of engineering a high efficiency, cost effective detection system, capable of meeting the above defined requirements in a variety of environmental situations. An array of NaI(Tl) detectors, arranged to form a 'line scan' along with a matching array of 'honeycomb' collimators provide a precisely defined field of view with minimal degradation of intrinsic detection efficiency and with significant scatter rejection. Scanning methods are adapted to net optimum detection efficiency of the combined system. In this work, and with differing detectors, a series of experimental demonstrations are performed that map system spatial performance and counting capability before actual spent fuel cask scans are performed. The data are evaluated to demonstrate the prompt ability to identify missing fuel rods or other content abnormalities. To also record and assess cask tampering, the cask is externally examined utilizing FTIR hyper spectral and other imaging/sensing approaches. This provides dated records and indications of external abnormalities (surface deposits, smears, contaminants, corrosion) attributable to normal degradation or to tampering. This paper will describe the actual gathering of data in both an experimental climate and from an actual spent fuel dry storage cask, and how an evaluation may be performed by an IAEA facility inspector attempting to draw an

  19. Photovoltage of Rods and Cones in the Macaque Retina

    NASA Astrophysics Data System (ADS)

    Schneeweis, David M.; Schnapf, Julie L.

    1995-05-01

    The kinetics, gain, and reliability of light responses of rod and cone photoreceptors are important determinants of overall visual sensitivity. In voltage recordings from photoreceptors in an intact primate retina, rods were found to be functionally isolated from each other, unlike the tightly coupled rods of cold-blooded vertebrates. Cones were observed to receive excitatory input from rods, which indicates that the cone pathway also processes rod signals. This input might be expected to degrade the spatial resolution of mesopic vision.

  20. Fuel Temperature Coefficient of Reactivity

    SciTech Connect

    Loewe, W.E.

    2001-07-31

    A method for measuring the fuel temperature coefficient of reactivity in a heterogeneous nuclear reactor is presented. The method, which is used during normal operation, requires that calibrated control rods be oscillated in a special way at a high reactor power level. The value of the fuel temperature coefficient of reactivity is found from the measured flux responses to these oscillations. Application of the method in a Savannah River reactor charged with natural uranium is discussed.

  1. Fuel flexible fuel injector

    SciTech Connect

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

    2015-02-03

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

  2. Rapid-L Operator-Free Fast Reactor Concept Without Any Control Rods

    SciTech Connect

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2003-07-15

    The 200-kW(electric) uranium-nitride-fueled lithium-cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for a lunar base power system. It is one of the variants of the RAPID (Refueling by All Pins Integrated Design) fast reactor concept, which enables quick and simplified refueling. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 2700 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 yr.Unique challenges in reactivity control systems design have been addressed in the RAPID-L concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt {sup 6}Li as a liquid poison instead of B{sub 4}C rods. In combination with LEMs, LIMs, and LRMs, RAPID-L can be operated without an operator. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, the RAPID-L reactor concept and its transient characteristics are presented.

  3. Simulation and operation of the EBR-2 automatic control rod drive system

    NASA Astrophysics Data System (ADS)

    Lehto, W. K.; Larson, H. A.; Dean, E. M.; Christensen, L. J.

    An automatic control rod drive system (ACRDS) installed at EBR-II produces shaped power transients from 40% to full reactor power at a linear ramp rate of 4 MWt/s. A digital computer and modified control rod drive provides this capability. Simulation and analysis of ACRDS experiments establish the safety envelope for reactor transient operation. Tailored transients are required as part of USDOE operational reliability testing program for prototypic fast reactor fuel cladding breach behavior studies. After initial EBR-II driver fuel testing and system checkout, test subassemblies were subjected to both slow and fast transients. In addition, the ACRDS is used for steady state operation and will be qualified to control power ascent from initial critical to full power.

  4. Preparation of high temperature gas-cooled reactor fuel element

    DOEpatents

    Bradley, Ronnie A.; Sease, John D.

    1976-01-01

    This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

  5. RADGEN: A radiation exchange factor generator for rod bundles

    SciTech Connect

    Rector, D.R.

    1987-10-01

    The RADGEN computer program has been developed at Pacific Northwest Laboratory (PNL) to generate input required for the thermal radiation models used in the COBRA-SFS (Spent Fuel Storage) computer program. The COBRA-SFS program uses radiation exchange factors to describe the net amount of energy transferred from each surface to every other surface in an enclosure. The RADGEN program generates radiation exchange factors for arrays of rods on a square or triangular pitch as well as open channel geometries. This report describes the input requirements for the RADGEN code, which may be executed in a batch or interactive mode, and outlines the solution procedure used to obtain the exchange factors. 4 refs., 25 figs., 13 tabs.

  6. Radiological characterization of spent control rod assemblies

    SciTech Connect

    Lepel, E.A.; Robertson, D.E.; Thomas, C.W.; Pratt, S.L.; Haggard, D.L.

    1995-10-01

    This document represents the final report of an ongoing study to provide radiological characterizations, classifications, and assessments in support of the decommissioning of nuclear power stations. This report describes the results of non-destructive and laboratory radionuclide measurements, as well as waste classification assessments, of BWR and PWR spent control rod assemblies. The radionuclide inventories of these spent control rods were determined by three separate methodologies, including (1) direct assay techniques, (2) calculational techniques, and (3) by sampling and laboratory radiochemical analyses. For the BWR control rod blade (CRB) and PWR burnable poison rod assembly (BPRA), {sup 60}Co and {sup 63}Ni, present in the stainless steel cladding, were the most abundant neutron activation products. The most abundant radionuclide in the PWR rod cluster control assembly (RCCA) was {sup 108m}Ag (130 yr halflife) produced in the Ag-In-Cd alloy used as the neutron poison. This radionuclide will be the dominant contributor to the gamma dose rate for many hundreds of years. The results of the direct assay methods agree very well ({+-}10%) with the sampling/radiochemical measurements. The results of the calculational methods agreed fairly well with the empirical measurements for the BPRA, but often varied by a factor of 5 to 10 for the CRB and the RCCA assemblies. If concentration averaging and encapsulation, as allowed by 10CFR61.55, is performed, then each of the entire control assemblies would be classified as Class C low-level radioactive waste.

  7. Passive electrical properties of rod outer segments

    PubMed Central

    Falk, Gertrude; Fatt, P.

    1968-01-01

    1. Measurements on a packed suspension of randomly oriented, dark-adapted frog rods at frequencies of 15 c/s-0·5 Mc/s indicate a behaviour similar to that of other biological materials. 2. Results are analysed on the assumption that the low-frequency limiting resistance is determined by current flowing in the suspending medium and that, of the rods, two thirds are oriented perpendicular to the applied field and one third parallel to it. Those parallel to the field are treated as non-conductors. 3. From the high-frequency limiting resistance the conductivity of the rod interior is calculated to vary linearly with the conductivity of the medium. The slope of the relation of internal to external conductivity is 0·50 with a limiting internal conductivity (at zero external) of 280 μmho/cm. 4. On the assumption that the suspension can be represented as a single-capacitance network, the characteristic frequency of impedance is used to calculate a capacitance for the rod surface of 1·54 μF/cm2. On the assumption of a distribution in properties of the suspension according to the theory of Bruggeman, the capacitance is calculated to have a value of about one half this. 5. At frequencies below 5 kc/s the impedance locus deviates from the curve describing the behaviour at higher frequencies. It is suggested that this may involve conduction in a thin layer extending along the surface of the rod. PMID:5685292

  8. Gelation And Mechanical Response of Patchy Rods

    NASA Astrophysics Data System (ADS)

    Kazem, Navid; Majidi, Carmel; Maloney, Craig

    We perform Brownian Dynamics simulations to study the gelation of suspensions of attractive, rod-like particles. We show that details of the particle-particle interactions can dramatically affect the dynamics of gelation and the structure and mechanics of the networks that form. If the attraction between the rods is perfectly smooth along their length, they will collapse into compact bundles. If the attraction is sufficiently corrugated or patchy, over time, a rigid space spanning network forms. We study the structure and mechanical properties of the networks that form as a function of the fraction of the surface that is allowed to bind. Surprisingly, the structural and mechanical properties are non-monotonic in the surface coverage. At low coverage, there are not a sufficient number of cross-linking sites to form networks. At high coverage, rods bundle and form disconnected clusters. At intermediate coverage, robust networks form. The elastic modulus and yield stress are both non-monotonic in the surface coverage. The stiffest and strongest networks show an essentially homogeneous deformation under strain with rods re-orienting along the extensional axis. Weaker, clumpy networks at high surface coverage exhibit relatively little re-orienting with strong non-affine deformation. These results suggest design strategies for tailoring surface interactions between rods to yield rigid networks with optimal properties. National Science Foundation and the Air Force Office of Scientific Research.

  9. Coiling of elastic rods on rigid substrates.

    PubMed

    Jawed, Mohammad K; Da, Fang; Joo, Jungseock; Grinspun, Eitan; Reis, Pedro M

    2014-10-14

    We investigate the deployment of a thin elastic rod onto a rigid substrate and study the resulting coiling patterns. In our approach, we combine precision model experiments, scaling analyses, and computer simulations toward developing predictive understanding of the coiling process. Both cases of deposition onto static and moving substrates are considered. We construct phase diagrams for the possible coiling patterns and characterize them as a function of the geometric and material properties of the rod, as well as the height and relative speeds of deployment. The modes selected and their characteristic length scales are found to arise from a complex interplay between gravitational, bending, and twisting energies of the rod, coupled to the geometric nonlinearities intrinsic to the large deformations. We give particular emphasis to the first sinusoidal mode of instability, which we find to be consistent with a Hopf bifurcation, and analyze the meandering wavelength and amplitude. Throughout, we systematically vary natural curvature of the rod as a control parameter, which has a qualitative and quantitative effect on the pattern formation, above a critical value that we determine. The universality conferred by the prominent role of geometry in the deformation modes of the rod suggests using the gained understanding as design guidelines, in the original applications that motivated the study. PMID:25267649

  10. Dynamic behavior of rod photoreceptor disks.

    PubMed

    Chen, Chunhe; Jiang, Yunhai; Koutalos, Yiannis

    2002-09-01

    Eukaryotic cells use membrane organelles, like the endoplasmic reticulum or the Golgi, to carry out different functions. Vertebrate rod photoreceptors use hundreds of membrane sacs (the disks) for the detection of light. We have used fluorescent tracers and single cell imaging to study the properties of rod photoreceptor disks. Labeling of intact rod photoreceptors with membrane markers and polar tracers revealed communication between intradiskal and extracellular space. Internalized tracers moved along the length of the rod outer segment, indicating communication between the disks as well. This communication involved the exchange of both membrane and aqueous phase and had a time constant in the order of minutes. The communication pathway uses approximately 2% of the available membrane disk area and does not allow the passage of molecules larger than 10 kDa. It was possible to load the intradiskal space with fluorescent Ca(2+) and pH dyes, which reported an intradiskal Ca(2+) concentration in the order of 1 microM and an acidic pH 6.5, both of them significantly different than intracellular and extracellular Ca(2+) concentrations and pH. The results suggest that the rod photoreceptor disks are not discrete, passive sacs but rather comprise an active cellular organelle. The communication between disks may be important for membrane remodeling as well as for providing access to the intradiskal space of the whole outer segment. PMID:12202366

  11. Single Rod Vibration in Axial Flow

    NASA Astrophysics Data System (ADS)

    Weichselbaum, Noah; Wang, Shengfu; Bardet, Philippe

    2013-11-01

    Fluid structure interaction of a single rod in axial flow is a coupled dynamical system present in many application including nuclear reactors, steam generators, and towed antenna arrays. Fluid-structure response can be quantified thanks to detailed experimental data where both structure and fluid responses are recorded. Such datum deepen understanding of the physics inherent to the system and provide high-dimensionality quantitative measurements to validate coupled structural and CFD codes with various level of complexity. In this work, single rods fixed on both ends in a concentric pipe, are subjected to an axial flow with Reynolds number based on hydraulic diameter of Re =4000. Rods of varying material stiffness and diameter are utilized in the experiment resulting in a range of dimensionless U between 0.5 and 1, where U = (ρA/EI)1/2uL. Experimental measurements of the velocity field around the rod are taken with PIV from time-resolved Nd:YLF laser and a high speed CMOS camera. Three-dimensional and temporal vibration and deflection of the rod is recorded with shadowgraphy utilizing two sets of pulsed high power LED and dedicated CMOS camera. Through integration of these two diagnostics, it is possible to reconstruct the full FSI domain providing unique validation data.

  12. Coiling of elastic rods on rigid substrates

    PubMed Central

    Jawed, Mohammad K.; Da, Fang; Joo, Jungseock; Grinspun, Eitan; Reis, Pedro M.

    2014-01-01

    We investigate the deployment of a thin elastic rod onto a rigid substrate and study the resulting coiling patterns. In our approach, we combine precision model experiments, scaling analyses, and computer simulations toward developing predictive understanding of the coiling process. Both cases of deposition onto static and moving substrates are considered. We construct phase diagrams for the possible coiling patterns and characterize them as a function of the geometric and material properties of the rod, as well as the height and relative speeds of deployment. The modes selected and their characteristic length scales are found to arise from a complex interplay between gravitational, bending, and twisting energies of the rod, coupled to the geometric nonlinearities intrinsic to the large deformations. We give particular emphasis to the first sinusoidal mode of instability, which we find to be consistent with a Hopf bifurcation, and analyze the meandering wavelength and amplitude. Throughout, we systematically vary natural curvature of the rod as a control parameter, which has a qualitative and quantitative effect on the pattern formation, above a critical value that we determine. The universality conferred by the prominent role of geometry in the deformation modes of the rod suggests using the gained understanding as design guidelines, in the original applications that motivated the study. PMID:25267649

  13. Criticality of spent reactor fuel

    SciTech Connect

    Harris, D.R.

    1987-01-01

    The storage capacity of spent reactor fuel pools can be greatly increased by consolidation. In this process, the fuel rods are removed from reactor fuel assemblies and are stored in close-packed arrays in a canister or skeleton. An earlier study examined criticality consideration for consolidation of Westinghouse fuel, assumed to be fresh, in canisters at the Millstone-2 spent-fuel pool and in the General Electric IF-300 shipping cask. The conclusions were that the fuel rods in the canister are so deficient in water that they are adequately subcritical, both in normal and in off-normal conditions. One potential accident, the water spill event, remained unresolved in the earlier study. A methodology is developed here for spent-fuel criticality and is applied to the water spill event. The methodology utilizes LEOPARD to compute few-group cross sections for the diffusion code PDQ7, which then is used to compute reactivity. These codes give results for fresh fuel that are in good agreement with KENO IV-NITAWL Monte Carlo results, which themselves are in good agreement with continuous energy Monte Carlo calculations. These methodologies are in reasonable agreement with critical measurements for undepleted fuel.

  14. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    SciTech Connect

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  15. Magnetic switch for reactor control rod

    DOEpatents

    Germer, John H.

    1986-01-01

    A magnetic reed switch assembly for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electromagnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

  16. Magnetic switch for reactor control rod. [LMFBR

    DOEpatents

    Germer, J.H.

    1982-09-30

    A magnetic reed switch assembly is described for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electro-magnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

  17. Studying rod photoreceptor development in zebrafish

    PubMed Central

    Morris, A.C.; Fadool, J.M.

    2009-01-01

    The zebrafish has rapidly become a favored model vertebrate organism, well suited for studies of developmental processes using large-scale genetic screens. In particular, zebrafish morphological and behavioral genetic screens have led to the identification of genes important for development of the retinal photoreceptors. This may help clarify the genetic mechanisms underlying human photoreceptor development and dysfunction in retinal diseases. In this review, we present the advantages of zebrafish as a vertebrate model organism, summarize retinal and photoreceptor cell development in zebrafish, with emphasis on the rod photoreceptors, and describe zebrafish visual behaviors that can be used for genetic screens. We then describe some of the photoreceptor cell mutants that have been isolated in morphological and behavioral screens and discuss the limitations of current screening methods for uncovering mutations that specifically affect rod function. Finally, we present some alternative strategies to target the rod developmental pathway in zebrafish. PMID:16199068

  18. Fragmentation of an axially impacted slender rod

    NASA Astrophysics Data System (ADS)

    Ji, W.; Waas, A. M.

    2010-02-01

    Motivated by experimental results on the dynamic buckling and fragmentation of a vertical column impacted by a falling mass, results from an analytical model for dynamic buckling which considers the dynamic interaction between the axial column deformation and the out-of-plane buckling displacements are used to interpret the fragmentation process and the resulting fragment lengths. It is shown that a critical time exists for the rod to undergo fragmentation. At this critical time, the rod deforms in a modulated pattern of waves, setting up the stage for the ensuing fragmentation as a result of induced large curvatures that exceed the critical bending strain of the rod material. The resulting fragment length distributions, which show two characteristics peaks at \\frac{\\lambda}{2} and \\frac{\\lambda}{4} , where λ is a characteristic half-wavelength, are found to compare favorably with the experimental results.

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

    SciTech Connect

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

    2013-05-01

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

  20. Effects of cooling time on a closed LWR fuel cycle

    SciTech Connect

    Arnold, R. P.; Forsberg, C. W.; Shwageraus, E.

    2012-07-01

    In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)

  1. HIGH STRENGTH CONTROL RODS FOR NEUTRONIC REACTORS

    DOEpatents

    Lustman, B.; Losco, E.F.; Cohen, I.

    1961-07-11

    Nuclear reactor control rods comprised of highly compressed and sintered finely divided metal alloy panticles and fine metal oxide panticles substantially uniformly distributed theretbrough are described. The metal alloy consists essentially of silver, indium, cadmium, tin, and aluminum, the amount of each being present in centain percentages by weight. The oxide particles are metal oxides of the metal alloy composition, the amount of oxygen being present in certain percentages by weight and all the oxygen present being substantially in the form of metal oxide. This control rod is characterized by its high strength and resistance to creep at elevated temperatures.

  2. Method for producing titanium aluminide weld rod

    DOEpatents

    Hansen, Jeffrey S.; Turner, Paul C.; Argetsinger, Edward R.

    1995-01-01

    A process for producing titanium aluminide weld rod comprising: attaching one end of a metal tube to a vacuum line; placing a means between said vacuum line and a junction of the metal tube to prevent powder from entering the vacuum line; inducing a vacuum within the tube; placing a mixture of titanium and aluminum powder in the tube and employing means to impact the powder in the tube to a filled tube; heating the tube in the vacuum at a temperature sufficient to initiate a high-temperature synthesis (SHS) reaction between the titanium and aluminum; and lowering the temperature to ambient temperature to obtain a intermetallic titanium aluminide alloy weld rod.

  3. Control rod drive for reactor shutdown

    DOEpatents

    McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

    1976-01-20

    A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

  4. Realizing actual feedback control of complex network

    NASA Astrophysics Data System (ADS)

    Tu, Chengyi; Cheng, Yuhua

    2014-06-01

    In this paper, we present the concept of feedbackability and how to identify the Minimum Feedbackability Set of an arbitrary complex directed network. Furthermore, we design an estimator and a feedback controller accessing one MFS to realize actual feedback control, i.e. control the system to our desired state according to the estimated system internal state from the output of estimator. Last but not least, we perform numerical simulations of a small linear time-invariant dynamics network and a real simple food network to verify the theoretical results. The framework presented here could make an arbitrary complex directed network realize actual feedback control and deepen our understanding of complex systems.

  5. An evaluation of the nuclear fuel performance code BISON

    SciTech Connect

    Perez, D. M.; Williamson, R. L.; Novascone, S. R.; Larson, T. K.; Hales, J. D.; Spencer, B. W.; Pastore, G.

    2013-07-01

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

  6. Experience in PWR and BWR mixed-oxide fuel management

    SciTech Connect

    Schlosser, G.J.; Krebs, W.; Urban, P. )

    1993-04-01

    Germany has adopted the strategy of a closed fuel cycle using reprocessing and recycling. The central issue today is plutonium recycling by the use of U-Pu mixed oxide (MOX) in pressurized water reactors (PWRs) and boiling water reactors (BWRs). The design of MOX fuel assemblies and fuel management in MOX-containing cores are strongly influenced by the nuclear properties of the plutonium isotopes. Optimized MOX fuel assembly designs for PWRs currently use up to three types of MOX fuel rods having different plutonium contents with natural uranium or uranium tailings as carrier material but without burnable absorbers. The MOX fuel assembly designs for BWRs use four to six rod types with different plutonium contents and Gd[sub 2]O[sub 3]/UO[sub 2] burnable absorber rods. Both the PWR and the BWR designs attain good burnup equivalence and compatibility with uranium fuel assemblies. High flexibility exists in the loading schemes relative to the position and number of MOX fuel assemblies in the reloads and in the core as a whole. The Siemens experience with MOX fuel assemblies is based on the insertion of 318 MOX fuel assemblies in eight PWRs and 168 in BWRs and pressurized heavy water reactors so far. The primary operating results include information on the cycle length, power distribution, reactivity coefficients, and control rod worth of cores containing MOX fuel assemblies.

  7. Some factors to consider in handling and storing spent fuel

    SciTech Connect

    Bailey, W.J.

    1985-11-01

    This report includes information from various studies performed under the Wet Storage Task of the Behavior of Spent Fuel in Storage Project of the Commercial Spent Fuel Management (CSFM) Program at Pacific Northwest Laboratory. Wet storage experience has been summarized earlier in several other reports. This report summarizes pertinent items noted during FY 1985 concerning recent developments in the handling and storage of spent fuel and associated considerations. The subjects discussed include recent publications, findings, and developments associated with: (1) storage of water reactor spent fuel in water pools, (2) extended-burnup fuel, (3) fuel assembly reconstitution and reinsertion, (4) rod consolidation, (5) variations in the US Nuclear Regulatory Commission's definition of failed fuel, (6) detection of failed fuel rods, and (7) extended integrity of spent fuel. A list of pertinent publications is included.

  8. Fuels for Sodium-cooled Fast Reactors: U.S. Perspective

    SciTech Connect

    Douglas C. Crawford; Douglas L. Porter; Steven L. Hayes

    2007-09-01

    The U.S. experience with mixed oxide, metal, and mixed carbide fuels is substantial, comprised of irradiation of over 50,000 MOX rods, over 130,000 metal rods, and 600 mixed carbide rods, in EBR-II and FFTF alone. All three types have all been demonstrated capable of fuel utilization at or above 200 GWd/MTHM. To varying degrees, life-limiting phenomena for each type have been identified and investigated, and there are no disqualifying safety-related fuel behaviors. All three fuel types appear capable of meeting SFR fuel requirements, with reliability of MOX and metal fuel well established. Improvements in irradiation performance of cladding and duct alloys has been a key development in moving these fuel designs toward higher-burnup potential. Selection of one fuel system over another will depend on circumstances particular to the application and on issues other than fuel performance, such as fabrication cost or overall system safety performance.

  9. Modeling and simulation performance of sucker rod beam pump

    SciTech Connect

    Aditsania, Annisa; Rahmawati, Silvy Dewi Sukarno, Pudjo; Soewono, Edy

    2015-09-30

    Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research.

  10. Modeling and simulation performance of sucker rod beam pump

    NASA Astrophysics Data System (ADS)

    Aditsania, Annisa; Rahmawati, Silvy Dewi; Sukarno, Pudjo; Soewono, Edy

    2015-09-01

    Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research.

  11. High Burnup Fuel Behavior Modeling

    SciTech Connect

    Jahingir, M.; Rand, R.; Stachowski, R.; Miles, B.; Kusagaya, K.

    2007-07-01

    This paper discusses the development and qualification of the PRIME03 code to address high burnup mechanisms and to improve uranium utilization in current and new reactor designs. Materials properties and behavioral models have been updated from previous thermal-mechanical codes to reflect the effects of burnup on fuel pellet thermal conductivity, Zircaloy creep, fuel pellet relocation, and fission gas release. These new models are based on results of in-pool and post irradiation examination (PIE) of commercial boiling water reactor (BWR) fuel rods at high burnup and results from international experimental programs. The new models incorporated into PRIME03 also address specific high burnup effects associated with formation of pellet rim porosity at high exposure. The PRIME03 code is qualified by comparison of predicted and measured fuel performance parameters for a large number of high, low, and moderate burnup test and commercial reactor rod. The extensive experimental qualification of the PRIME03 prediction capabilities confirms that it is a reliable best-estimate predictor of fuel rod thermal-mechanical performance over a wide range of design and operating conditions. (authors)

  12. 50 CFR 253.16 - Actual cost.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 50 Wildlife and Fisheries 9 2011-10-01 2011-10-01 false Actual cost. 253.16 Section 253.16 Wildlife and Fisheries NATIONAL MARINE FISHERIES SERVICE, NATIONAL OCEANIC AND ATMOSPHERIC ADMINISTRATION, DEPARTMENT OF COMMERCE AID TO FISHERIES FISHERIES ASSISTANCE PROGRAMS Fisheries Finance Program §...

  13. 50 CFR 253.16 - Actual cost.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 50 Wildlife and Fisheries 11 2013-10-01 2013-10-01 false Actual cost. 253.16 Section 253.16 Wildlife and Fisheries NATIONAL MARINE FISHERIES SERVICE, NATIONAL OCEANIC AND ATMOSPHERIC ADMINISTRATION, DEPARTMENT OF COMMERCE AID TO FISHERIES FISHERIES ASSISTANCE PROGRAMS Fisheries Finance Program §...

  14. Humanistic Education and Self-Actualization Theory.

    ERIC Educational Resources Information Center

    Farmer, Rod

    1984-01-01

    Stresses the need for theoretical justification for the development of humanistic education programs in today's schools. Explores Abraham Maslow's hierarchy of needs and theory of self-actualization. Argues that Maslow's theory may be the best available for educators concerned with educating the whole child. (JHZ)

  15. Children's Rights and Self-Actualization Theory.

    ERIC Educational Resources Information Center

    Farmer, Rod

    1982-01-01

    Educators need to seriously reflect upon the concept of children's rights. Though the idea of children's rights has been debated numerous times, the idea remains vague and shapeless; however, Maslow's theory of self-actualization can provide the children's rights idea with a needed theoretical framework. (Author)

  16. Culture Studies and Self-Actualization Theory.

    ERIC Educational Resources Information Center

    Farmer, Rod

    1983-01-01

    True citizenship education is impossible unless students develop the habit of intelligently evaluating cultures. Abraham Maslow's theory of self-actualization, a theory of innate human needs and of human motivation, is a nonethnocentric tool which can be used by teachers and students to help them understand other cultures. (SR)

  17. Group Counseling for Self-Actualization.

    ERIC Educational Resources Information Center

    Streich, William H.; Keeler, Douglas J.

    Self-concept, creativity, growth orientation, an integrated value system, and receptiveness to new experiences are considered to be crucial variables to the self-actualization process. A regular, year-long group counseling program was conducted with 85 randomly selected gifted secondary students in the Farmington, Connecticut Public Schools. A…

  18. Racial Discrimination in Occupations: Perceived and Actual.

    ERIC Educational Resources Information Center

    Turner, Castellano B.; Turner, Barbara F.

    The relationship between the actual representation of Blacks in certain occupations and individual perceptions of the occupational opportunity structure were examined. A scale which rated the degree of perceived discrimination against Blacks in 21 occupations was administered to 75 black male, 70 black female, 1,429 white male and 1,457 white…

  19. Developing Human Resources through Actualizing Human Potential

    ERIC Educational Resources Information Center

    Clarken, Rodney H.

    2012-01-01

    The key to human resource development is in actualizing individual and collective thinking, feeling and choosing potentials related to our minds, hearts and wills respectively. These capacities and faculties must be balanced and regulated according to the standards of truth, love and justice for individual, community and institutional development,…

  20. Fuel cell stack monitoring and system control

    DOEpatents

    Keskula, Donald H.; Doan, Tien M.; Clingerman, Bruce J.

    2005-01-25

    A control method for monitoring a fuel cell stack in a fuel cell system in which the actual voltage and actual current from the fuel cell stack are monitored. A preestablished relationship between voltage and current over the operating range of the fuel cell is established. A variance value between the actual measured voltage and the expected voltage magnitude for a given actual measured current is calculated and compared with a predetermined allowable variance. An output is generated if the calculated variance value exceeds the predetermined variance. The predetermined voltage-current for the fuel cell is symbolized as a polarization curve at given operating conditions of the fuel cell. Other polarization curves may be generated and used for fuel cell stack monitoring based on different operating pressures, temperatures, hydrogen quantities.

  1. Whiteheadian Actual Entitities and String Theory

    NASA Astrophysics Data System (ADS)

    Bracken, Joseph A.

    2012-06-01

    In the philosophy of Alfred North Whitehead, the ultimate units of reality are actual entities, momentary self-constituting subjects of experience which are too small to be sensibly perceived. Their combination into "societies" with a "common element of form" produces the organisms and inanimate things of ordinary sense experience. According to the proponents of string theory, tiny vibrating strings are the ultimate constituents of physical reality which in harmonious combination yield perceptible entities at the macroscopic level of physical reality. Given that the number of Whiteheadian actual entities and of individual strings within string theory are beyond reckoning at any given moment, could they be two ways to describe the same non-verifiable foundational reality? For example, if one could establish that the "superject" or objective pattern of self- constitution of an actual entity vibrates at a specific frequency, its affinity with the individual strings of string theory would be striking. Likewise, if one were to claim that the size and complexity of Whiteheadian 'societies" require different space-time parameters for the dynamic interrelationship of constituent actual entities, would that at least partially account for the assumption of 10 or even 26 instead of just 3 dimensions within string theory? The overall conclusion of this article is that, if a suitably revised understanding of Whiteheadian metaphysics were seen as compatible with the philosophical implications of string theory, their combination into a single world view would strengthen the plausibility of both schemes taken separately. Key words: actual entities, subject/superjects, vibrating strings, structured fields of activity, multi-dimensional physical reality.

  2. Piston rod seal for a Stirling engine

    SciTech Connect

    Shapiro, W.

    1984-01-31

    In a piston rod seal for a Stirling engine, a hydrostatic bearing and differential pressure regulating valve are utilized to provide for a low pressure differential across a rubbing seal between the hydrogen and oil so as to reduce wear on the seal. 3 figs.

  3. Dark Current and Photocurrent in Retinal Rods

    PubMed Central

    Hagins, W. A.; Penn, R. D.; Yoshikami, S.

    1970-01-01

    The interstitial voltages, currents, and resistances of the receptor layer of the isolated rat retina have been investigated with arrays of micropipette electrodes inserted under direct visual observation by infrared microscopy. In darkness a steady current flows inward through the plasma membrane of the rod outer segments. It is balanced by equal outward current distributed along the remainder of each rod. Flashes of light produce a photocurrent which transiently reduces the dark current with a waveform resembling the PII and a-wave components of the electroretinogram. The photocurrent is produced by a local action of light within 12 μm of its point of absorption in the outer segments. The quantum current gain of the photocurrent is greater than 106. The electrical space constant of rat rods is greater than 25 μm, so that the electrical effects of the photocurrent are large enough at the rod synapses to permit single absorbed photons to be detected by the visual system. The photocurrent is apparently the primary sensory consequence of light absorption by rhodopsin. ImagesFigure 3Figure 8Figure 14 PMID:5439318

  4. CONTROL ROD ALLOY CONTAINING NOBLE METAL ADDITIONS

    DOEpatents

    Anderson, W.K.; Ray, W.E.

    1960-05-01

    Silver-base alloys suitable for use in the fabrication of control rods for neutronic reactors are given. The alloy consists of from 0.5 wt.% to about 1.5 wt.% of a noble metal of platinum, ruthenium, rhodium, osmium, or palladium, up to 10 wt.% of cadmium, from 2 to 20 wt.% indium, the balance being silver.

  5. Image analysis for measuring rod network properties

    NASA Astrophysics Data System (ADS)

    Kim, Dongjae; Choi, Jungkyu; Nam, Jaewook

    2015-12-01

    In recent years, metallic nanowires have been attracting significant attention as next-generation flexible transparent conductive films. The performance of films depends on the network structure created by nanowires. Gaining an understanding of their structure, such as connectivity, coverage, and alignment of nanowires, requires the knowledge of individual nanowires inside the microscopic images taken from the film. Although nanowires are flexible up to a certain extent, they are usually depicted as rigid rods in many analysis and computational studies. Herein, we propose a simple and straightforward algorithm based on the filtering in the frequency domain for detecting the rod-shape objects inside binary images. The proposed algorithm uses a specially designed filter in the frequency domain to detect image segments, namely, the connected components aligned in a certain direction. Those components are post-processed to be combined under a given merging rule in a single rod object. In this study, the microscopic properties of the rod networks relevant to the analysis of nanowire networks were measured for investigating the opto-electric performance of transparent conductive films and their alignment distribution, length distribution, and area fraction. To verify and find the optimum parameters for the proposed algorithm, numerical experiments were performed on synthetic images with predefined properties. By selecting proper parameters, the algorithm was used to investigate silver nanowire transparent conductive films fabricated by the dip coating method.

  6. A Cambrian origin for vertebrate rods

    PubMed Central

    Asteriti, Sabrina; Grillner, Sten; Cangiano, Lorenzo

    2015-01-01

    Vertebrates acquired dim-light vision when an ancestral cone evolved into the rod photoreceptor at an unknown stage preceding the last common ancestor of extant jawed vertebrates (∼420 million years ago Ma). The jawless lampreys provide a unique opportunity to constrain the timing of this advance, as their line diverged ∼505 Ma and later displayed high-morphological stability. We recorded with patch electrodes the inner segment photovoltages and with suction electrodes the outer segment photocurrents of Lampetra fluviatilis retinal photoreceptors. Several key functional features of jawed vertebrate rods are present in their phylogenetically homologous photoreceptors in lamprey: crucially, the efficient amplification of the effect of single photons, measured by multiple parameters, and the flow of rod signals into cones. These results make convergent evolution in the jawless and jawed vertebrate lines unlikely and indicate an early origin of rods, implying strong selective pressure toward dim-light vision in Cambrian ecosystems. DOI: http://dx.doi.org/10.7554/eLife.07166.001 PMID:26095697

  7. Monodisperse hard rods in external potentials

    NASA Astrophysics Data System (ADS)

    Bakhti, Benaoumeur; Karbach, Michael; Maass, Philipp; Müller, Gerhard

    2015-10-01

    We consider linear arrays of cells of volume Vc populated by monodisperse rods of size σ Vc,σ =1 ,2 ,... , subject to hardcore exclusion interaction. Each rod experiences a position-dependent external potential. In one application we also examine effects of contact forces between rods. We employ two distinct methods of exact analysis with complementary strengths and different limits of spatial resolution to calculate profiles of pressure and density on mesoscopic and microscopic length scales at thermal equilibrium. One method uses density functionals and the other statistically interacting vacancy particles. The applications worked out include gravity, power-law traps, and hard walls. We identify oscillations in the profiles on a microscopic length scale and show how they are systematically averaged out on a well-defined mesoscopic length scale to establish full consistency between the two approaches. The continuum limit, realized as Vc→0 ,σ →∞ at nonzero and finite σ Vc , connects our highest-resolution results with known exact results for monodisperse rods in a continuum. We also compare the pressure profiles obtained from density functionals with the average microscopic pressure profiles derived from the pair distribution function.

  8. Piston rod seal for a Stirling engine

    DOEpatents

    Shapiro, Wilbur

    1984-01-01

    In a piston rod seal for a Stirling engine, a hydrostatic bearing and differential pressure regulating valve are utilized to provide for a low pressure differential across a rubbing seal between the hydrogen and oil so as to reduce wear on the seal.

  9. Drop Ejection From an Oscillating Rod

    NASA Technical Reports Server (NTRS)

    Wilkes, E. D.; Basaran, O. A.

    1999-01-01

    The dynamics of a drop of a Newtonian liquid that is pendant from or sessile on a solid rod that is forced to undergo time-periodic oscillations along its axis is studied theoretically. The free boundary problem governing the time evolution of the shape of the drop and the flow field inside it is solved by a method of lines using a finite element algorithm incorporating an adaptive mesh. When the forcing amplitude is small, the drop approaches a limit cycle at large times and undergoes steady oscillations thereafter. However, drop breakup is the consequence if the forcing amplitude exceeds a critical value. Over a wide range of amplitudes above this critical value, drop ejection from the rod occurs during the second oscillation period from the commencement of rod motion. Remarkably, the shape of the interface at breakup and the volume of the primary drop formed are insensitive to changes in forcing amplitude. The interface shape at times close to and at breakup is a multi-valued function of distance measured along the rod axis and hence cannot be described by recently popularized one-dimensional approximations. The computations show that drop ejection occurs without the formation of a long neck. Therefore, this method of drop formation holds promise of preventing formation of undesirable satellite droplets.

  10. Stimulus-evoked outer segment changes in rod photoreceptors

    NASA Astrophysics Data System (ADS)

    Zhao, Xiaohui; Thapa, Damber; Wang, Benquan; Lu, Yiming; Gai, Shaoyan; Yao, Xincheng

    2016-06-01

    Rod-dominated transient retinal phototropism (TRP) has been recently observed in freshly isolated mouse and frog retinas. Comparative confocal microscopy and optical coherence tomography revealed that the TRP was predominantly elicited from the rod outer segment (OS). However, the biophysical mechanism of rod OS dynamics is still unknown. Mouse and frog retinal slices, which displayed a cross-section of retinal photoreceptors and other functional layers, were used to test the effect of light stimulation on rod OSs. Time-lapse microscopy revealed stimulus-evoked conformational changes of rod OSs. In the center of the stimulated region, the length of the rod OS shrunk, while in the peripheral region, the rod OS swung toward the center region. Our experimental observation and theoretical analysis suggest that the TRP may reflect unbalanced rod disc-shape changes due to localized visible light stimulation.

  11. Reprogramming of adult rod photoreceptors prevents retinal degeneration

    PubMed Central

    Montana, Cynthia L.; Kolesnikov, Alexander V.; Shen, Susan Q.; Myers, Connie A.; Kefalov, Vladimir J.; Corbo, Joseph C.

    2013-01-01

    A prime goal of regenerative medicine is to direct cell fates in a therapeutically useful manner. Retinitis pigmentosa is one of the most common degenerative diseases of the eye and is associated with early rod photoreceptor death followed by secondary cone degeneration. We hypothesized that converting adult rods into cones, via knockdown of the rod photoreceptor determinant Nrl, could make the cells resistant to the effects of mutations in rod-specific genes, thereby preventing secondary cone loss. To test this idea, we engineered a tamoxifen-inducible allele of Nrl to acutely inactivate the gene in adult rods. This manipulation resulted in reprogramming of rods into cells with a variety of cone-like molecular, histologic, and functional properties. Moreover, reprogramming of adult rods achieved cellular and functional rescue of retinal degeneration in a mouse model of retinitis pigmentosa. These findings suggest that elimination of Nrl in adult rods may represent a unique therapy for retinal degeneration. PMID:23319618

  12. 5. DETAIL OF ROD MILL BASE AND CONVEYOR BELT SUPPORT, ...

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

    5. DETAIL OF ROD MILL BASE AND CONVEYOR BELT SUPPORT, EAST VIEW. - Vanadium Corporation of America (VCA) Naturita Mill, Grinding Rod Mill, 3 miles Northwest of Naturita, between Highway 141 & San Miguel River, Naturita, Montrose County, CO

  13. 32 CFR 989.21 - Record of decision (ROD).

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... the signator. A ROD (40 CFR 1505.2) is a concise public document stating what an agency's decision is... ENVIRONMENTAL IMPACT ANALYSIS PROCESS (EIAP) § 989.21 Record of decision (ROD). (a) The proponent and the...

  14. 32 CFR 989.21 - Record of decision (ROD).

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... the signator. A ROD (40 CFR 1505.2) is a concise public document stating what an agency's decision is... ENVIRONMENTAL IMPACT ANALYSIS PROCESS (EIAP) § 989.21 Record of decision (ROD). (a) The proponent and the...

  15. 11. Detail showing stay rods and westerly Samson post; view ...

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

    11. Detail showing stay rods and westerly Samson post; view to west. Note slackness of stay rods while draw-span is closed. - Summer Street Bridge, Spanning Reserved Channel, Boston, Suffolk County, MA

  16. COBRA-SFS (Spent-Fuel Storage) thermal-hydraulic analyses of the CASTOR-1C and REA 2023 BWR storage casks containing consolidated spent fuel

    SciTech Connect

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.

    1986-12-01

    Consolidation of spent nuclear fuel rods is being considered as one option for more efficient and compact storage of reactor spent fuel assemblies. In this concept, rods from two disassembled spent fuel assemblies will be consolidated in a space originally intended to store a single unconsolidated assembly. The thermal performance of consolidated fuel rods in dry storage, especially in multiassembly storage systems, is one of the major issues that must be addressed prior to implementation. In this study, Pacific Northwest Laboratory researchers performed thermal-hydraulic analyses for both the REA 2023 cask and the CASTOR-1C cask containing either unconsolidated or consolidated BWR spent fuel assemblies. The objective was to determine the effect of consolidating spent fuel assemblies on the temperature distributions within both types of casks. Two major conclusions resulted from this study. First, a lumping technique (combining rods and flow channels), which reduces the number of computational nodes required to model complex multiassembly geometries, could be used for both unconsolidated and consolidated rods with negligible effect on prediction accuracies. Second, with a relatively high thermal conductivity backfill gas (e.g., helium), the predicted peak fuel rod temperature in a canister of consolidated rods generating the same amount of heat as an unconsolidated assembly is essentially the same as the peak temperature in the unconsolidated assembly. In contrast, with a relatively low thermal conductivity backfill gas (e.g., nitrogen), the opposite is true and the predicted peak temperature in a consolidated canister is significantly higher than in an unconsolidated assembly. Therefore, when rods are consolidated, selection of the backfill gas is important in maintaining peak rod temperatures below allowable values for rods with relatively high decay heat generation rates.

  17. Microfluidic fuel cells

    NASA Astrophysics Data System (ADS)

    Kjeang, Erik

    Microfluidic fuel cell architectures are presented in this thesis. This work represents the mechanical and microfluidic portion of a microfluidic biofuel cell project. While the microfluidic fuel cells developed here are targeted to eventual integration with biocatalysts, the contributions of this thesis have more general applicability. The cell architectures are developed and evaluated based on conventional non-biological electrocatalysts. The fuel cells employ co-laminar flow of fuel and oxidant streams that do not require a membrane for physical separation, and comprise carbon or gold electrodes compatible with most enzyme immobilization schemes developed to date. The demonstrated microfluidic fuel cell architectures include the following: a single cell with planar gold electrodes and a grooved channel architecture that accommodates gaseous product evolution while preventing crossover effects; a single cell with planar carbon electrodes based on graphite rods; a three-dimensional hexagonal array cell based on multiple graphite rod electrodes with unique scale-up opportunities; a single cell with porous carbon electrodes that provides enhanced power output mainly attributed to the increased active area; a single cell with flow-through porous carbon electrodes that provides improved performance and overall energy conversion efficiency; and a single cell with flow-through porous gold electrodes with similar capabilities and reduced ohmic resistance. As compared to previous results, the microfluidic fuel cells developed in this work show improved fuel cell performance (both in terms of power density and efficiency). In addition, this dissertation includes the development of an integrated electrochemical velocimetry approach for microfluidic devices, and a computational modeling study of strategic enzyme patterning for microfluidic biofuel cells with consecutive reactions.

  18. 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.

  19. 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.

  20. An Evaluation of the PBF LOFT Lead Rod Test Results Concerning Surface Thermocouple Perturbation Effects

    SciTech Connect

    M. L. Carboneau E. L. Tolman

    1980-02-08

    The purpose of the Power Burst Facility Loss of Fluid Test (PBF LOFT) Lead Rod (LLR) Test program was to provide experimental data to characterize the mechanical behavior of LOFT type nuclear fuel rods under loss of coolant accident (LOCA) conditions, simulating the test conditions expected for the LOFT Power Ascension (L2) Test series. Although the LLR tests were not explicitly designed to evaluate cladding surface thermocouple perturbation effects, comparison of the Linear Variable Differential Transformer (LVDT) data for rods instrumented with and without cladding thermocouples provided pertinent information concerning the effects of cladding thermocouples on the time to DNB and time to quench data. Documentation and review of this data is presented in the following report. It will be shown that most of the LLR data indicate that the cladding surface thermocouples did not enhance the rewetting characteristics of the rods they are attached to, even though other evidence shows that the surface clad thermocouples did quench early. Finally, in order to accurately interpret and understand the limitations of the LVDT instrumentation, upon which thermocouple perturbation effects were evaluated, an analysis of the LVDT data as well as a review of the atypical response events that occurred during the LLR tests are presented in appendices to this document.

  1. Viscosity of concentrated suspensions of sphere/rod mixtures

    SciTech Connect

    Mor, R.; Gottlieb, M.; Graham, A.L.; Mondy, L.A.

    1996-05-01

    This paper discusses the viscosity of concentrated suspensions of sphere/rod mixtures by adopting the Thomas relations for spheres and Milliken`s for randomly oriented rods with aspect ratio of 20. The relative viscosity of a mixed suspension may now be calculated for any combination of rods (of aspect ratio 20) and spheres.

  2. CONTROL ROD DRIVE MECHANISM FOR A NUCLEAR REACTOR

    DOEpatents

    Hawke, B.C.; Liederbach, F.J.; Lones, W.

    1963-05-14

    A lead-screw-type control rod drive featuring an electric motor and a fluid motor arranged to provide a selectably alternative driving means is described. The electric motor serves to drive the control rod slowly during normal operation, while the fluid motor, assisted by an automatic declutching of the electric motor, affords high-speed rod insertion during a scram. (AEC)

  3. 21 CFR 888.3020 - Intramedullary fixation rod.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 8 2014-04-01 2014-04-01 false Intramedullary fixation rod. 888.3020 Section 888.3020 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES ORTHOPEDIC DEVICES Prosthetic Devices § 888.3020 Intramedullary fixation rod. (a) Identification. An intramedullary fixation rod...

  4. 21 CFR 888.3020 - Intramedullary fixation rod.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 8 2013-04-01 2013-04-01 false Intramedullary fixation rod. 888.3020 Section 888.3020 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES ORTHOPEDIC DEVICES Prosthetic Devices § 888.3020 Intramedullary fixation rod. (a) Identification. An intramedullary fixation rod...

  5. Carbon Inverse Opal Rods for Nonenzymatic Cholesterol Detection.

    PubMed

    Zhong, Qifeng; Xie, Zhuoying; Ding, Haibo; Zhu, Cun; Yang, Zixue; Gu, Zhongze

    2015-11-18

    Carbon inverse opal rods made from silica photonic crystal rods are used for nonenzymatic cholesterol sensing. The characteristic reflection peak originating from the physical periodic structure works as sensing signals for quantitatively estimating cholesterol concentrations. Carbon inverse opal rods work both in cholesterol standard solutions and human serum. They are suitable for practical use in clinical diagnose. PMID:26415111

  6. Rod Has High Tensile Strength And Low Thermal Expansion

    NASA Technical Reports Server (NTRS)

    Smith, D. E.; Everton, R. L.; Howe, E.; O'Malley, M.

    1996-01-01

    Thoriated tungsten extension rod fabricated to replace stainless-steel extension rod attached to linear variable-differential transformer in gap-measuring gauge. Threads formed on end of rod by machining with special fixtures and carefully chosen combination of speeds and feeds.

  7. Measurement of the Speed of Sound in a Metal Rod.

    ERIC Educational Resources Information Center

    Mak, Se-yuen; Ng, Yee-kong; Wu, Kam-wah

    2000-01-01

    Suggests two improved methods to measure the speed of sound in a metal rod. One employs a fast timer to measure the time required for a compression pulse to travel along the rod from end to end, and a second uses a microphone to measure the frequency of the fundamental mode of a freely suspending singing rod. (Author/ASK)

  8. Technical and regulatory review of the Rover nuclear fuel process for use on Fort St. Vrain fuel

    SciTech Connect

    Hertzler, T. )

    1993-02-01

    This report describes the results of an analysis for processing and final disposal of Fort St. Vrain (FSV) irradiated fuel in Rover-type equipment or technologies. This analysis includes an evaluation of the current Rover equipment status and the applicability of this technology in processing FSV fuel. The analyses are based on the physical characteristics of the FSV fuel and processing capabilities of the Rover equipment. Alternate FSV fuel disposal options are also considered including fuel-rod removal from the block, disposal of the empty block, or disposal of the entire fuel-containing block. The results of these analyses document that the current Rover hardware is not operable for any purpose, and any effort to restart this hardware will require extensive modifications and re-evaluation. However, various aspects of the Rover technology, such as the successful fluid-bed burner design, can be applied with modification to FSV fuel processing. The current regulatory climate and technical knowledge are not adequately defined to allow a complete analysis and conclusion with respect to the disposal of intact fuel blocks with or without the fuel rods removed. The primary unknowns include the various aspects of fuel-rod removal from the block, concentration of radionuclides remaining in the graphite block after rod removal, and acceptability of carbon in the form of graphite in a high level waste repository.

  9. Axial grading of inert matrix fuels

    SciTech Connect

    Recktenwald, G. D.; Deinert, M. R.

    2012-07-01

    Burning actinides in an inert matrix fuel to 750 MWd/kg IHM results in a significant reduction in transuranic isotopes. However, achieving this level of burnup in a standard light water reactor would require residence times that are twice that of uranium dioxide fuels. The reactivity of an inert matrix assembly at the end of life is less than 1/3 of its beginning of life reactivity leading to undesirable radial and axial power peaking in the reactor core. Here we show that axial grading of the inert matrix fuel rods can reduce peaking significantly. Monte Carlo simulations are used to model the assembly level power distributions in both ungraded and graded fuel rods. The results show that an axial grading of uranium dioxide and inert matrix fuels with erbium can reduces power peaking by more than 50% in the axial direction. The reduction in power peaking enables the core to operate at significantly higher power. (authors)

  10. The Actual Apollo 13 Prime Crew

    NASA Technical Reports Server (NTRS)

    1970-01-01

    The actual Apollo 13 lunar landing mission prime crew from left to right are: Commander, James A. Lovell Jr., Command Module pilot, John L. Swigert Jr.and Lunar Module pilot, Fred W. Haise Jr. The original Command Module pilot for this mission was Thomas 'Ken' Mattingly Jr. but due to exposure to German measles he was replaced by his backup, Command Module pilot, John L. 'Jack' Swigert Jr.

  11. Simulation of spatial fuel assay using HANARO neutron beam

    PubMed

    Lee; Chang; Lee; Kim

    2000-10-01

    A sensitivity simulation of neutron tomography was performed for the analysis of the spatial distribution of nuclear materials in the HANARO fuel rod. The internal distribution of the nuclear materials in the fuel rod is very important for the increase of the safety and economics of fuel burnup in the reactor. The neutron radiography facility installed at HANARO will be used for the spatial fuel analysis with a real-time image processing system. Monte Carlo simulation was performed to study the feasibility and sensitivity of the HANARO neutron beam for the spatial fuel assay and to find the optimum conditions for neutron detection. From the sensitivity simulation, the location of the nuclear materials in the rod was evident as expected. PMID:11003495

  12. Cavitation phenomena in mechanical heart valves: studied by using a physical impinging rod system.

    PubMed

    Lo, Chi-Wen; Chen, Sheng-Fu; Li, Chi-Pei; Lu, Po-Chien

    2010-10-01

    When studying mechanical heart valve cavitation, a physical model allows direct flow field and pressure measurements that are difficult to perform with actual valves, as well as separate testing of water hammer and squeeze flow effects. Movable rods of 5 and 10 mm diameter impinged same-sized stationary rods to simulate squeeze flow. A 24 mm piston within a tube simulated water hammer. Adding a 5 mm stationary rod within the tube generated both effects simultaneously. Charged-coupled device (CCD) laser displacement sensors, strobe lighting technique, laser Doppler velocimetry (LDV), particle image velocimetry (PIV) and high fidelity piezoelectric pressure transducers measured impact velocities, cavitation images, squeeze flow velocities, vortices, and pressure changes at impact, respectively. The movable rods created cavitation at critical impact velocities of 1.6 and 1.2 m/s; squeeze flow velocities were 2.8 and 4.64 m/s. The isolated water hammer created cavitation at 1.3 m/s piston speed. The combined piston and stationary rod created cavitation at an impact speed of 0.9 m/s and squeeze flow of 3.2 m/s. These results show squeeze flow alone caused cavitation, notably at lower impact velocity as contact area increased. Water hammer alone also caused cavitation with faster displacement. Both effects together were additive. The pressure change at the vortex center was only 150 mmHg, which cannot generate the magnitude of pressure drop required for cavitation bubble formation. Cavitation occurred at 3-5 m/s squeeze flow, significantly different from the 14 m/s derived by Bernoulli's equation; the temporal acceleration of unsteady flow requires further study. PMID:20490686

  13. Fatigue Analysis of the Piston Rod in a Kaplan Turbine Based on Crack Propagation under Unsteady Hydraulic Loads

    NASA Astrophysics Data System (ADS)

    Liu, X.; Y Luo, Y.; Wang, Z. W.

    2014-03-01

    As an important component of the blade-control system in Kaplan turbines, piston rods are subjected to fluctuating forces transferred by the turbines blades from hydraulic pressure oscillations. Damage due to unsteady hydraulic loads might generate unexpected down time and high repair cost. In one running hydropower plant, the fracture failure of the piston rod was found twice at the same location. With the transient dynamic analysis, the retainer ring structure of the piston rod existed a relative high stress concentration. This predicted position of the stress concentration agreed well with the actual fracture position in the plant. However, the local strain approach was not able to explain why this position broke frequently. Since traditional structural fatigue analyses use a local stress strain approach to assess structural integrity, do not consider the effect of flaws which can significantly degrade structural life. Using linear elastic fracture mechanism (LEFM) approaches that include the effect of flaws is becoming common practice in many industries. In this research, a case involving a small semi-ellipse crack was taken into account at the stress concentration area, crack growth progress was calculated by FEM. The relationship between crack length and remaining life was obtained. The crack propagation path approximately agreed with the actual fracture section. The results showed that presence of the crack had significantly changed the local stress and strain distributions of the piston rod compared with non-flaw assumption.

  14. Rod Ejection Accident by the Coupled System Code ATHLET-QUABOX/CUBBOX

    NASA Astrophysics Data System (ADS)

    Perin, Yann; Velkov, Kiril; Pasichnyk, Igor; Langenbuch, Siegfried

    The paper considers a Rod Ejection Accident (REA) which has been calculated by the coupled-code system ATHLET-QUABOX/CUBBOX. For the present study, a MOX/UOX mixed core loading was developed on the basis of a generic PWR. The results are particularly focused on the fuel enthalpy rise which is the main safety criterion for such transient. A parametric REA study has been performed, showing the influence of some important thermal-hydraulic and neutron-physical parameters. Simulations have been performed using realistic or artificially decreased delayed neutron fractions for two different core states (HZP and 30% of the nominal power). Effective fuel rod temperature influence (i.e. Doppler coefficient) has been studied by using different correlations (0.5/0.5 weighting factors or the typical TDoppler = 0.7 TSurface + 0.3 TCenter) or by changing the fuel gap conductance. It is shown that the maximum enthalpy (and enthalpy increase) does not always appear in the affected fuel assembly but can also appear in the neighboring ones. This result is a direct consequence of the burn up dependence of the enthalpy. The paper also considers the case of local delayed neutron parameters and briefly describes the future REA studies foreseen at GRS such as an investigation of quantitative uncertainty propagation from the nuclear data to the transient behavior.

  15. Early User Experience with BISON Fuel Performance Code

    SciTech Connect

    D. M. Perez

    2012-08-01

    Three Fuel Modeling Exercise II (FUMEX II) LWR fuel irradiation experiments were simulated and analyzed using the fuel performance code BISON to demonstrate code utility for modeling of the LWR fuel performance. Comparisons were made against the BISON results and the experimental data for the three assessment cases. The assessment cases reported within this report include IFA-597.3 Rod 8, Riso AN3 and Riso AN4.

  16. Mie resonance in the arrays of dielectric rods in air

    NASA Astrophysics Data System (ADS)

    Dalal, Reena; Kalra, Yogita; Sinha, R. K.

    2015-09-01

    Mie resonance in square arrays of dielectric rods has been reported. Arrays in square lattice of dielectric rods with very high permittivity in air have been considered. Light of transverse electric mode has been launched on the square array of cylindrical dielectric rods. Mie resonance of first two orders has been observed in the dielectric rods, due to which electric and magnetic dipoles are generated in the rods. Thus, electric resonance and magnetic resonance at different frequencies has been observed with material of high value of permittivity.

  17. Diffusion of a nanowire rod through an obstacle field

    NASA Astrophysics Data System (ADS)

    Kasimov, Dror; Admon, Tamir; Roichman, Yael

    2016-05-01

    We report the experimental realization of a rod diffusing in a two-dimensional obstacle field following the single rod dynamics. We use a silver nanowire as our rod and two types of obstacles: repelling light beams and polymer pillars. We study the effect of hydrodynamic interactions on the transport of the rod, comparing both experimental realizations and recent simulations. We propose a framework for analyzing the transport through such systems, and we predict a new superdiffusive regime of rod transport at high obstacle concentration and short times.

  18. Following the equilibria of slender elastic rods

    NASA Astrophysics Data System (ADS)

    Lazarus, Arnaud; Miller, James; Reis, Pedro

    2012-02-01

    We present a novel continuation method to characterize and quantify the equilibria of elastic rods under large geometrically nonlinear displacements and rotations. To describe the kinematics we exploit the synthetic power and computational efficiency of quaternions. The energetics of bending, stretching and torsion are all taken into account to derive the equilibrium equations which we solve using an asymptotic numerical continuation method. This provides access to the full set of analytical equilibrium branches (stable and unstable), a.k.a bifurcation diagrams. This is in contrast with the individual solution points attained by classic energy minimization or predictor-corrector techniques. We challenge our numerics for the specific problem of an extremely twisted naturally curved rod and perform a detailed comparison against a precision desktop-scale experiments. The quantification of the underlying 3D buckling instabilities and the characterization of the resulting complex configurations are in excellent agreement between numerics and experiments.

  19. Dynamics of gas-fluidized granular rods.

    PubMed

    Daniels, L J; Park, Y; Lubensky, T C; Durian, D J

    2009-04-01

    We study a quasi-two-dimensional monolayer of granular rods fluidized by a spatially and temporally homogeneous upflow of air. By tracking the position and orientation of the particles, we characterize the dynamics of the system with sufficient resolution to observe ballistic motion at the shortest time scales. Particle anisotropy gives rise to dynamical anisotropy and superdiffusive dynamics parallel to the rod's long axis, causing the parallel and perpendicular mean-square displacements to become diffusive on different time scales. The distributions of free times and free paths between collisions deviate from exponential behavior, underscoring the nonthermal character of the particle motion. The dynamics show evidence of rotational-translational coupling similar to that of an anisotropic Brownian particle. We model rotational-translational coupling in the single-particle dynamics with a modified Langevin model using nonthermal noise sources. This suggests a phenomenological approach to thinking about collections of self-propelling particles in terms of enhanced memory effects. PMID:19518218

  20. On the Vortex Sound from Rotating Rods

    NASA Technical Reports Server (NTRS)

    Yudin, E. Y.

    1947-01-01

    The motion of different bodies imersed in liquid or gaseous media is accompanied by characteristic sound which is excited by the formation of unstable surfaces of separation behind the body, usually disintegrating into a system of discrete vortices(such as the Karman vortex street due to the flow about an infintely long rod, etc.).In the noise from fans,pumps,and similar machtnery, vortexnQif3eI?Yequently predominates. The purpose of this work is to elucidate certain questions of the dependence ofthis sound upon the aerodynamic parameters and the tip speed of the rotating rods,or blades. Although scme material is given below,insufficientto calculate the first rough approximation to the solution of this question,such as the mechanics of vortex formation,never the less certain conclusions maybe found of practical application for the reduction of noise from rotating blades.

  1. Hamiltonian formulations and symmetries in rod mechanics

    SciTech Connect

    Dichmann, D.J.; Li, Yiwei; Maddocks, J.H.

    1996-12-31

    This article provides a survey of contemporary rod mechanics, including both dynamic and static theories. Much of what we discuss is regarded as classic material within the mechanics community, but the objective here is to provide a self-contained account accessible to workers interested in modelling DNA. We also describe a number of recent results and computations involving rod mechanics that have been obtained by our group at the University of Maryland. This work was largely motivated by applications to modelling DNA, but our approach reflects a background of research in continuum mechanics. In particular, we emphasize the role that Hamiltonian formulations and symmetries play in the effective computation of special solutions, conservation laws of dynamics and integrals of statics. 41 refs., 10 figs.

  2. Theory of thin-walled rods

    NASA Technical Reports Server (NTRS)

    Goldenveizer, A L

    1951-01-01

    Starting with the Love equations for bending of extensible shells, "principal stress states" are sought for a thin-walled rod of arbitrary but open cross section. Principal stress states exclude those local states arising from end conditions which damp out with distance from the ends. It is found that for rods of intermediate length, long enough to avoid local bending at a support, and short enough that elementary torsion and bending are not the most significant stress states, four principal states exist. Three of these states are associated with the planar distribution of axial stress and are equivalent to the engineering theory of extension and bending of solid sections. The fourth state resembles that which has been called in the literature "bending stress due to torsional", except that cross sections are permitted to bend and the shear along the center line of the cross section is permitted to differ from zero.

  3. [Rod of Asclepius. Symbol of medicine].

    PubMed

    Young, Pablo; Finn, Bárbara C; Bruetman, Julio E; Cesaro Gelos, Jorge; Trimarchi, Hernán

    2013-09-01

    Symbolism is one of the most archaic forms of human thoughts. Symbol derives from the Latin word symbolum, and the latter from the Greek symbolon or symballo, which means "I coincide, I make matches". The Medicine symbol represents a whole series of historical and ethical values. Asclepius Rod with one serpent entwined, has traditionally been the symbol of scientific medicine. In a misconception that has lasted 500 years, the Caduceus of Hermes, entwined by two serpents and with two wings, has been considered the symbol of Medicine. However, the Caduceus is the current symbol of Commerce. Asclepius Rod and the Caduceus of Hermes represent two professions, Medicine and Commerce that, in ethical practice, should not be mixed. Physicians should be aware of their real emblem, its historical origin and meaning. PMID:24522424

  4. Statistical properties of a folded elastic rod

    NASA Astrophysics Data System (ADS)

    Bayart, Elsa; Deboeuf, Stéphanie; Boué, Laurent; Corson, Francis; Boudaoud, Arezki; Adda-Bedia, Mokhtar

    2010-03-01

    A large variety of elastic structures naturally seem to be confined into environments too small to accommodate them; the geometry of folded structures span a wide range of length-scales. The elastic properties of these confined systems are further constrained by self-avoidance as well as by the dimensionality of both structures and container. To mimic crumpled paper, we devised an experimental setup to study the packing of a dimensional elastic object in 2D geometries: an elastic rod is folded at the center of a circular Hele-Shaw cell by a centripetal force. The initial configuration of the rod and the acceleration of the rotating disk allow to span different final folded configurations while the final rotation speed controls the packing intensity. Using image analysis we measure geometrical and mechanical properties of the folded configurations, focusing on length, curvature and energy distributions.

  5. Fluid Dynamics in Sucker Rod Pumps

    SciTech Connect

    Cutler, R.P.; Mansure, A.J.

    1999-01-14

    Sucker rod pumps are installed in approximately 90% of all oil wells in the U.S. Although they have been widely used for decades, there are many issues regarding the fluid dynamics of the pump that have not been fully investigated. A project was conducted at Sandia National Laboratories to develop unimproved understanding of the fluid dynamics inside a sucker rod pump. A mathematical flow model was developed to predict pressures in any pump component or an entire pump under single-phase fluid and pumping conditions. Laboratory flow tests were conducted on instrumented individual pump components and on a complete pump to verify and refine the model. The mathematical model was then converted to a Visual Basic program to allow easy input of fluid, geometry and pump parameters and to generate output plots. Examples of issues affecting pump performance investigated with the model include the effects of viscosity, surface roughness, valve design details, plunger and valve pressure differentials, and pumping rate.

  6. A survey of laser lightning rod techniques

    NASA Technical Reports Server (NTRS)

    Barnes, Arnold A., Jr.; Berthel, Robert O.

    1991-01-01

    The work done to create a laser lightning rod (LLR) is discussed. Some ongoing research which has the potential for achieving an operational laser lightning rod for use in the protection of missile launch sites, launch vehicles, and other property is discussed. Because of the ease with which a laser beam can be steered into any cloud overhead, an LLR could be used to ascertain if there exists enough charge in the clouds to discharge to the ground as triggered lightning. This leads to the possibility of using LLRs to test clouds prior to launching missiles through the clouds or prior to flying aircraft through the clouds. LLRs could also be used to probe and discharge clouds before or during any hazardous ground operations. Thus, an operational LLR may be able to both detect such sub-critical electrical fields and effectively neutralize them.

  7. Evaluation of ANF fuel failures in oyster creek

    SciTech Connect

    Howe, T.M.; Van Swam, L.F.; Piascik, T.G.; Spence, P.A.

    1988-01-01

    During the refueling outrage following cycle-10 operations of Oyster Creek nuclear generating station, fuel sipping identified 47 failed Advance Nuclear Fuels (ANF) fuel assemblies. The failed fuel was an unpressurized 8 x 8 design manufactured by ANF prior to 1980. Subsequent inspection of 46 of these 47 assemblies with the ANF ULTRATEST ultrasonic testing system indicated 104 either failed of suspect fuel rods in 44 assemblies. Two of the assemblies were identified as being sound. Selected fuel rods were removed from three of the assemblies and inspected both visually and with an eddycurrent coil. An evaluation has been performed to determine the cause of the failures. The failures were primarily the result of pellet/cladding interaction (PCI). Detailed analyses of several of the failed fuel rods were performed with ANF's fuel rod modeling code RAMPX2. RAMPX2 includes several state-of-the-art models, including a model describing the formation of fission product deposits called coins on the inside surface of the cladding, a model that accounts for axial PCI, and a trapped fuel stack model. The analyses provided an explanation for the failures.

  8. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: II - Rod Bowing Effect on Boiling Transition under Transient Conditions

    NASA Astrophysics Data System (ADS)

    Liu, Wei; Tamai, Hidesada; Kureta, Masatoshi; Ohnuki, Akira; Akimoto, Hajime

    A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we describe the critical power characteristics in a 37-rod tight-lattice bundle with rod bowing under transient states. It is observed that transient Boiling Transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle, which is same as that under steady state. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with a modified TRAC-BFI code, where Japan Atomic Energy Agency (JAEA) newest critical power correlation is implemented for the BT judgement. The code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time conservatively. Traditional quasi-steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight-lattice bundle with rod bowing.

  9. Rod Driven Frequency Entrainment and Resonance Phenomena.

    PubMed

    Salchow, Christina; Strohmeier, Daniel; Klee, Sascha; Jannek, Dunja; Schiecke, Karin; Witte, Herbert; Nehorai, Arye; Haueisen, Jens

    2016-01-01

    A controversy exists on photic driving in the human visual cortex evoked by intermittent photic stimulation. Frequency entrainment and resonance phenomena are reported for frequencies higher than 12 Hz in some studies while missing in others. We hypothesized that this might be due to different experimental conditions, since both high and low intensity light stimulation were used. However, most studies do not report radiometric measurements, which makes it impossible to categorize the stimulation according to photopic, mesopic, and scotopic vision. Low intensity light stimulation might lead to scotopic vision, where rod perception dominates. In this study, we investigated photic driving for rod-dominated visual input under scotopic conditions. Twelve healthy volunteers were stimulated with low intensity light flashes at 20 stimulation frequencies, leading to rod activation only. The frequencies were multiples of the individual alpha frequency (α) of each volunteer in the range from 0.40 to 2.30(∗)α. Three hundred and six-channel whole head magnetoencephalography recordings were analyzed in time, frequency, and spatiotemporal domains with the Topographic Matching Pursuit algorithm. We found resonance phenomena and frequency entrainment for stimulations at or close to the individual alpha frequency (0.90-1.10(∗)α) and half of the alpha frequency (0.40-0.55(∗)α). No signs of resonance and frequency entrainment phenomena were revealed around 2.00(∗)α. Instead, on-responses at the beginning and off-responses at the end of each stimulation train were observed for the first time in a photic driving experiment at frequencies of 1.30-2.30(∗)α, indicating that the flicker fusion threshold was reached. All results, the resonance and entrainment as well as the fusion effects, provide evidence for rod-dominated photic driving in the visual cortex. PMID:27588002

  10. On the perfect hexagonal packing of rods

    NASA Astrophysics Data System (ADS)

    Starostin, E. L.

    2006-04-01

    In most cases the hexagonal packing of fibrous structures or rods extremizes the energy of interaction between strands. If the strands are not straight, then it is still possible to form a perfect hexatic bundle. Conditions under which the perfect hexagonal packing of curved tubular structures may exist are formulated. Particular attention is given to closed or cycled arrangements of the rods like in the DNA toroids and spools. The closure or return constraints of the bundle result in an allowable group of automorphisms of the cross-sectional hexagonal lattice. The structure of this group is explored. Examples of open helical-like and closed toroidal-like bundles are presented. An expression for the elastic energy of a perfectly packed bundle of thin elastic rods is derived. The energy accounts for both the bending and torsional stiffnesses of the rods. It is shown that equilibria of the bundle correspond to solutions of a variational problem formulated for the curve representing the axis of the bundle. The functional involves a function of the squared curvature under the constraints on the total torsion and the length. The Euler-Lagrange equations are obtained in terms of curvature and torsion and due to the existence of the first integrals the problem is reduced to the quadrature. The three-dimensional shape of the bundle may be readily reconstructed by integration of the Ilyukhin-type equations in special cylindrical coordinates. The results are of universal nature and are applicable to various fibrous structures, in particular, to intramolecular liquid crystals formed by DNA condensed in toroids or packed inside the viral capsids. International Workshop on Biopolymers: Thermodynamics, Kinetics and Mechanics of DNA, RNA and Proteins, 30.05.2005-3.06.2005, The Abdus Salam International Centre for Theoretical Physics (ICTP), Trieste, Italy.

  11. Rod Driven Frequency Entrainment and Resonance Phenomena

    PubMed Central

    Salchow, Christina; Strohmeier, Daniel; Klee, Sascha; Jannek, Dunja; Schiecke, Karin; Witte, Herbert; Nehorai, Arye; Haueisen, Jens

    2016-01-01

    A controversy exists on photic driving in the human visual cortex evoked by intermittent photic stimulation. Frequency entrainment and resonance phenomena are reported for frequencies higher than 12 Hz in some studies while missing in others. We hypothesized that this might be due to different experimental conditions, since both high and low intensity light stimulation were used. However, most studies do not report radiometric measurements, which makes it impossible to categorize the stimulation according to photopic, mesopic, and scotopic vision. Low intensity light stimulation might lead to scotopic vision, where rod perception dominates. In this study, we investigated photic driving for rod-dominated visual input under scotopic conditions. Twelve healthy volunteers were stimulated with low intensity light flashes at 20 stimulation frequencies, leading to rod activation only. The frequencies were multiples of the individual alpha frequency (α) of each volunteer in the range from 0.40 to 2.30∗α. Three hundred and six-channel whole head magnetoencephalography recordings were analyzed in time, frequency, and spatiotemporal domains with the Topographic Matching Pursuit algorithm. We found resonance phenomena and frequency entrainment for stimulations at or close to the individual alpha frequency (0.90–1.10∗α) and half of the alpha frequency (0.40–0.55∗α). No signs of resonance and frequency entrainment phenomena were revealed around 2.00∗α. Instead, on-responses at the beginning and off-responses at the end of each stimulation train were observed for the first time in a photic driving experiment at frequencies of 1.30–2.30∗α, indicating that the flicker fusion threshold was reached. All results, the resonance and entrainment as well as the fusion effects, provide evidence for rod-dominated photic driving in the visual cortex. PMID:27588002

  12. Measurements of lightning rod responses to nearby strikes

    NASA Astrophysics Data System (ADS)

    Moore, C. B.; Aulich, G. D.; Rison, W.

    2000-05-01

    Following Benjamin Franklin's invention of the lightning rod, based on his discovery that electrified objects could be discharged by approaching them with a metal needle in hand, conventional lightning rods in the U.S. have had sharp tips. In recent years, the role of the sharp tip in causing a lightning rod to act as a strike receptor has been questioned leading to experiments in which pairs of various sharp-tipped and blunt rods have been exposed beneath thunderclouds to determine the better strike receptor. After seven years of tests, none of the sharp Franklin rods or of the so-called “early streamer emitters” has been struck, but 12 blunt rods with tip diameters ranging from 12.7 mm to 25.4 mm have taken strikes. Our field experiments and our analyses indicate that the strike-reception probabilities of Franklin's rods are greatly increased when their tips are made moderately blunt.

  13. ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

    DOEpatents

    Young, J.N.

    1957-08-20

    An electromagnetic device for moving an object in a linear path by increments is described. The device is specifically adapted for moving a neutron absorbing control rod into and out of the core of a reactor and consists essentially of an extension member made of magnetic material connected to one end of the control rod and mechanically flexible to grip the walls of a sleeve member when flexed, a magnetic sleeve member coaxial with and slidable between limit stops along the flexible extension, electromagnetic coils substantially centrally located with respect to the flexible extension to flex the extension member into gripping engagement with the sleeve member when ener gized, moving electromagnets at each end of the sleeve to attract the sleeve when energized, and a second gripping electromagnet positioned along the flexible extension at a distance from the previously mentioned electromagnets for gripping the extension member when energized. In use, the second gripping electromagnet is deenergized, the first gripping electromagnet is energized to fix the extension member in the sleeve, and one of the moving electromagnets is energized to attract the sleeve member toward it, thereby moving the control rod.

  14. Incorporation of squalene into rod outer segments

    SciTech Connect

    Keller, R.K.; Fliesler, S.J. )

    1990-08-15

    We have reported previously that squalene is the major radiolabeled nonsaponifiable lipid product derived from ({sup 3}H)acetate in short term incubations of frog retinas. In the present study, we demonstrate that newly synthesized squalene is incorporated into rod outer segments under similar in vitro conditions. We show further that squalene is an endogenous constituent of frog rod outer segment membranes; its concentration is approximately 9.5 nmol/mumol of phospholipid or about 9% of the level of cholesterol. Pulse-chase experiments with radiolabeled precursors revealed no metabolism of outer segment squalene to sterols in up to 20 h of chase. Taken together with our previous absolute rate studies, these results suggest that most, if not all, of the squalene synthesized by the frog retina is transported to rod outer segments. Synthesis of protein is not required for squalene transport since puromycin had no effect on squalene incorporation into outer segments. Conversely, inhibition of isoprenoid synthesis with mevinolin had no effect on the incorporation of opsin into the outer segment. These latter results support the conclusion that the de novo synthesis and subsequent intracellular trafficking of opsin and isoprenoid lipids destined for the outer segment occur via independent mechanisms.

  15. Interfacial phenomena in hard-rod fluids

    NASA Astrophysics Data System (ADS)

    Shundyak, K. Y.

    2004-05-01

    This thesis addresses questions of interfacial ordering in hard-rod fluids at coexistence of the isotropic and nematic phases and in their contact with simple model substrates. It is organized as follows. Chapter II provides some background information about the relation between the statistical mechanical and thermodynamical level of descriptions of bulk hard-rod fluids, as well as introduces the asymptotically exact Onsager model, and some basic facts of interfacial thermodynamics. Chapter III represents studies of the simplest free IN interface in a fluid of monodisperse Onsager hard rods. For the analysis of this system we develop an efficient perturbative method to determine the (biaxial) one-particle distribution function in inhomogeneous systems. Studies of the free planar isotropic-nematic interfaces are continued in Chapter IV, where they are considered in binary mixtures of hard rods. For sufficiently different particle shapes the bulk phase diagrams of these mixtures exhibit a triple point, where an isotropic (I) phase coexists with two nematic phases (N1 and N2) of different composition. For all explored mixtures we find that upon approach of the triple point the IN2 interface shows complete wetting by an intervening N1 film. We compute the surface tension of isotropic-nematic interfaces, and find a remarkable increase with fractionation. These studies are complemented by an analysis of bulk phase behavior and interfacial properties of nonadditive binary mixtures of thin and thick hard rods in Chapter V. The formulation of this model was motivated by recent experiments in the group of Fraden, who explored the phase behavior of a mixture of viruses with different effective diameters. In our model, species of the same types are considered as interacting with the hard-core repulsive potential, whereas the excluded volume for dissimilar rods is taken to be larger (smaller) then for the pure hard rods. Such a nonadditivity enhances (reduces) fractionation at

  16. Air resistance measurements on actual airplane parts

    NASA Technical Reports Server (NTRS)

    Weiselsberger, C

    1923-01-01

    For the calculation of the parasite resistance of an airplane, a knowledge of the resistance of the individual structural and accessory parts is necessary. The most reliable basis for this is given by tests with actual airplane parts at airspeeds which occur in practice. The data given here relate to the landing gear of a Siemanms-Schuckert DI airplane; the landing gear of a 'Luftfahrzeug-Gesellschaft' airplane (type Roland Dlla); landing gear of a 'Flugzeugbau Friedrichshafen' G airplane; a machine gun, and the exhaust manifold of a 269 HP engine.

  17. Evaluation of irradiated fuel during RIA simulation tests. Final report

    SciTech Connect

    Montgomery, R.O.; Rashid, Y.R.

    1996-08-01

    A critical assessment of the RIA-simulation experiments performed to date on previously irradiated test rods is presented. Included in this assessment are the SPERT-CDC, the NSRR, and the CABRI REP Na experimental programs. Information was collected describing the base irradiation, test rod characterization, and test procedures and conditions. The representativeness of the test rods and test conditions to anticipated LWR RIA accident conditions was evaluated using analysis results from fuel behavior and three-dimensional spatial kinetics simulations. It was shown that the pulse characteristics and coolant conditions are significantly different from those anticipated in an LWR-Furthermore, the unrepresentative test conditions were found to exaggerate the mechanisms that caused cladding failure. The data review identified several test rods which contained unusual cladding damage incurred prior to the RIA-simulation test that produced the observed failures. The mechanisms responsible for the observed test rod failures have been shown to result from processes that have a second order effect of burnup. A correlation with burnup could not be appropriately established for the fuel enthalpy at failure. However, the successful test rods can be used to construct a conservative region of success for fuel rod behavior during an RIA event.

  18. Prototypical Rod Consolidation Demonstration Project. Phase 3, Final report: Volume 1, Cold checkout test report, Book 2

    SciTech Connect

    Not Available

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 2 discusses the following topics: Fuel Rod Extraction System Test Results and Analysis Reports and Clamping Table Test Results and Analysis Reports.

  19. Aviation turbine fuel properties and their trends

    NASA Technical Reports Server (NTRS)

    Friedman, R.

    1981-01-01

    Fuel property values and their trends were studied through a review of a recognized, wide ranging sample population from actual fuel inspection data. A total of 676 fuel samples of Jet A aviation turbine fuel were compiled over an eleven year period. Results indicate that most fuel samples have one to three near-specification properties, the most common being aromatics, smoke point, and freezing point.

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

    SciTech Connect

    Bates, J.M.

    1986-01-01

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

  1. ADF/cofilin-actin rods in neurodegenerative diseases

    PubMed Central

    Bamburg, J.R.; Bernstein, B.W.; Davis, R.C.; Flynn, K.C.; Goldsbury, C.; Jensen, J.R.; Maloney, M.T.; Marsden, I.T.; Minamide, L.S.; Pak, C.W.; Shaw, A.E.; Whiteman, I.; Wiggan, O.

    2015-01-01

    Dephosphorylation (activation) of cofilin, an actin binding protein, is stimulated by initiators of neuronal dysfunction and degeneration including oxidative stress, excitotoxic glutamate, ischemia, and soluble forms of β-amyloid peptide (Aβ). Hyperactive cofilin forms rod-shaped cofilin-saturated actin filament bundles (rods). Other proteins are recruited to rods but are not necessary for rod formation. Neuronal cytoplasmic rods accumulate within neurites where they disrupt synaptic function and are a likely cause of synaptic loss without neuronal loss, as occurs early in dementias. Different rod-inducing stimuli target distinct neuronal populations within the hippocampus. Rods form rapidly, often in tandem arrays, in response to stress. They accumulate phosphorylated tau that immunostains for epitopes present in “striated neuropil threads,” characteristic of tau pathology in Alzheimer disease (AD) brain. Thus, rods might aid in further tau modifications or assembly into paired helical filaments, the major component of neurofibrillary tangles (NFTs). Rods can occlude neurites and block vesicle transport. Some rod-inducing treatments cause an increase in secreted Aβ. Thus rods may mediate the loss of synapses, production of excess Aβ, and formation of NFTs, all of the pathological hallmarks of AD. Cofilin-actin rods also form within the nucleus of heat-shocked neurons and are cleared from cells expressing wild type huntingtin protein but not in cells expressing mutant or silenced huntingtin, suggesting a role for nuclear rods in Huntington disease (HD). As an early event in the neurodegenerative cascade, rod formation is an ideal target for therapeutic intervention that might be useful in treatment of many different neurological diseases. PMID:20088812

  2. Fuel Burn Estimation Using Real Track Data

    NASA Technical Reports Server (NTRS)

    Chatterji, Gano B.

    2011-01-01

    A procedure for estimating fuel burned based on actual flight track data, and drag and fuel-flow models is described. The procedure consists of estimating aircraft and wind states, lift, drag and thrust. Fuel-flow for jet aircraft is determined in terms of thrust, true airspeed and altitude as prescribed by the Base of Aircraft Data fuel-flow model. This paper provides a theoretical foundation for computing fuel-flow with most of the information derived from actual flight data. The procedure does not require an explicit model of thrust and calibrated airspeed/Mach profile which are typically needed for trajectory synthesis. To validate the fuel computation method, flight test data provided by the Federal Aviation Administration were processed. Results from this method show that fuel consumed can be estimated within 1% of the actual fuel consumed in the flight test. Next, fuel consumption was estimated with simplified lift and thrust models. Results show negligible difference with respect to the full model without simplifications. An iterative takeoff weight estimation procedure is described for estimating fuel consumption, when takeoff weight is unavailable, and for establishing fuel consumption uncertainty bounds. Finally, the suitability of using radar-based position information for fuel estimation is examined. It is shown that fuel usage could be estimated within 5.4% of the actual value using positions reported in the Airline Situation Display to Industry data with simplified models and iterative takeoff weight computation.

  3. Between a Map and a Data Rod

    NASA Technical Reports Server (NTRS)

    Teng, William; Rui, Hualan; Strub, Richard; Vollmer, Bruce

    2015-01-01

    A Digital Divide has long stood between how NASA and other satellite-derived data are typically archived (time-step arrays or maps) and how hydrology and other point-time series oriented communities prefer to access those data. In essence, the desired method of data access is orthogonal to the way the data are archived. Our approach to bridging the Divide is part of a larger NASA-supported data rods project to enhance access to and use of NASA and other data by the Consortium of Universities for the Advancement of Hydrologic Science, Inc. (CUAHSI) Hydrologic Information System (HIS) and the larger hydrology community. Our main objective was to determine a way to reorganize data that is optimal for these communities. Two related objectives were to optimally reorganize data in a way that (1) is operational and fits in and leverages the existing Goddard Earth Sciences Data and Information Services Center (GES DISC) operational environment and (2) addresses the scaling up of data sets available as time series from those archived at the GES DISC to potentially include those from other Earth Observing System Data and Information System (EOSDIS) data archives. Through several prototype efforts and lessons learned, we arrived at a non-database solution that satisfied our objectivesconstraints. We describe, in this presentation, how we implemented the operational production of pre-generated data rods and, considering the tradeoffs between length of time series (or number of time steps), resources needed, and performance, how we implemented the operational production of on-the-fly (virtual) data rods. For the virtual data rods, we leveraged a number of existing resources, including the NASA Giovanni Cache and NetCDF Operators (NCO) and used data cubes processed in parallel. Our current benchmark performance for virtual generation of data rods is about a years worth of time series for hourly data (9,000 time steps) in 90 seconds. Our approach is a specific implementation of

  4. Between a Map and a Data Rod

    NASA Astrophysics Data System (ADS)

    Teng, W. L.; Rui, H.; Strub, R. F.; Vollmer, B.

    2015-12-01

    A "Digital Divide" has long stood between how NASA and other satellite-derived data are typically archived (time-step arrays or "maps") and how hydrology and other point-time series oriented communities prefer to access those data. In essence, the desired method of data access is orthogonal to the way the data are archived. Our approach to bridging the Divide is part of a larger NASA-supported "data rods" project to enhance access to and use of NASA and other data by the Consortium of Universities for the Advancement of Hydrologic Science, Inc. (CUAHSI) Hydrologic Information System (HIS) and the larger hydrology community. Our main objective was to determine a way to reorganize data that is optimal for these communities. Two related objectives were to optimally reorganize data in a way that (1) is operational and fits in and leverages the existing Goddard Earth Sciences Data and Information Services Center (GES DISC) operational environment and (2) addresses the scaling up of data sets available as time series from those archived at the GES DISC to potentially include those from other Earth Observing System Data and Information System (EOSDIS) data archives. Through several prototype efforts and lessons learned, we arrived at a non-database solution that satisfied our objectives/constraints. We describe, in this presentation, how we implemented the operational production of pre-generated data rods and, considering the tradeoffs between length of time series (or number of time steps), resources needed, and performance, how we implemented the operational production of on-the-fly ("virtual") data rods. For the virtual data rods, we leveraged a number of existing resources, including the NASA Giovanni Cache and NetCDF Operators (NCO) and used data cubes processed in parallel. Our current benchmark performance for virtual generation of data rods is about a year's worth of time series for hourly data (~9,000 time steps) in ~90 seconds. Our approach is a specific

  5. FUEL HANDLING MECHANISM

    DOEpatents

    Koch, L.J.; Hutter, E.

    1960-02-01

    A remotely operable handling device specifically adapted for the handling of vertically disposed fuel rods in a nuclear reactor was developed. The device consists essentially of an elongated tubular member having a gripping device at the lower end of the pivoted jaw type adapted to grip an enlarged head on the upper end of the workpiece. The device includes a sensing element which engages the enlarged head and is displaced to remotely indicate when the workpiece is in the proper position to be engaged by the jaws.

  6. PWR FLECHT SEASET 21-rod bundle flow blockage task data and analysis report. NRC/EPRI/Westinghouse Report No. 11. Appendices K-P

    SciTech Connect

    Loftus, M.J.; Hochreiter, L.E.; Lee, N.; McGuire, M.F.; Wenzel, A.H.; Valkovic, M.M.

    1982-09-01

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests.

  7. The ATLAS tile calorimeter ROD injector and multiplexer board

    NASA Astrophysics Data System (ADS)

    Valero, A.; Castillo, V.; Ferrer, A.; González, V.; Hernández, Y.; Higón, E.; Sanchís, E.; Solans, C.; Torres, J.; Valls, J. A.

    2011-02-01

    The ATLAS Tile Calorimeter is a sampling detector composed by cells made of iron-scintillator tiles. The calorimeter cell signals are digitized in the front-end electronics and transmitted to the Read-Out Drivers (RODs) at the first level trigger rate. The ROD receives triggered data from up to 9856 channels and provides the energy, phase and quality factor of the signals to the second level trigger. The back-end electronics is divided into four partitions containing eight RODs each. Therefore, a total of 32 RODs are used to process and transmit the data of the TileCal detector. In order to emulate the detector signals in the production and commissioning of ROD modules a board called ROD Injector and Multiplexer Board (RIMBO) was designed. In this paper, the RIMBO main functional blocks, PCB design and the different operation modes are described. It is described the crucial role of the board within the TileCal ROD test-bench in order to emulate the front-end electronics during the validation of ROD boards as well as during the evaluation of the ROD signal reconstruction algorithms. Finally, qualification and performance results for the injection operation mode obtained during the Tile Calorimeter ROD production tests are presented.

  8. Calibration of a fuel relocation model in BISON

    SciTech Connect

    Swiler, L. P.; Williamson, R. L.; Perez, D. M.

    2013-07-01

    We demonstrate parameter calibration in the context of the BISON nuclear fuels performance analysis code. Specifically, we present the calibration of a parameter governing fuel relocation: the power level at which the relocation model is activated. This relocation activation parameter is a critical value in obtaining reasonable comparison with fuel centerline temperature measurements. It also is the subject of some debate in terms of the optimal values. We show that the optimal value does vary across the calibration to individual rods. We also demonstrate an aggregated calibration, where we calibrate to observations from six rods. (authors)

  9. ANALYSIS AND EXAMINATION OF MOX FUEL FROM NONPROLIFERATION PROGRAMS

    SciTech Connect

    McCoy, Kevin; Machut, Dr McLean; Morris, Robert Noel; Blanpain, Patrick; Hemrick, James Gordon

    2013-01-01

    The U.S. Department of Energy has decided to dispose of a portion of the nation s surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. Four lead assemblies were manufactured and irradiated to a maximum fuel rod burnup of 47.3 MWd/kg heavy metal. This was the first commercial irradiation of MOX fuel with a 240Pu/239Pu ratio of less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. The performance of the rods was analyzed with AREVA s next-generation GALILEO code. The results of the analysis confirmed that the fuel rods had performed safely and predictably, and that GALILEO is applicable to MOX fuel with a low 240Pu/239Pu ratio as well as to standard MOX. The results are presented and compared to the GALILEO database. In addition, the fuel cladding was tested to confirm that traces of gallium in the fuel pellets had not affected the mechanical properties of the cladding. The irradiated cladding was found to remain ductile at both room temperature and 350 C for both the axial and circumferential directions.

  10. 7 CFR 1437.101 - Actual production history.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 7 Agriculture 10 2012-01-01 2012-01-01 false Actual production history. 1437.101 Section 1437.101... Determining Yield Coverage Using Actual Production History § 1437.101 Actual production history. Actual production history (APH) is the unit's record of crop yield by crop year for the APH base period. The...

  11. 7 CFR 1437.101 - Actual production history.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 7 Agriculture 10 2014-01-01 2014-01-01 false Actual production history. 1437.101 Section 1437.101... Determining Yield Coverage Using Actual Production History § 1437.101 Actual production history. Actual production history (APH) is the unit's record of crop yield by crop year for the APH base period. The...

  12. 7 CFR 1437.101 - Actual production history.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 7 Agriculture 10 2013-01-01 2013-01-01 false Actual production history. 1437.101 Section 1437.101... Determining Yield Coverage Using Actual Production History § 1437.101 Actual production history. Actual production history (APH) is the unit's record of crop yield by crop year for the APH base period. The...

  13. 7 CFR 1437.101 - Actual production history.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 7 Agriculture 10 2011-01-01 2011-01-01 false Actual production history. 1437.101 Section 1437.101... Determining Yield Coverage Using Actual Production History § 1437.101 Actual production history. Actual production history (APH) is the unit's record of crop yield by crop year for the APH base period. The...

  14. 7 CFR 1437.101 - Actual production history.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Actual production history. 1437.101 Section 1437.101... Determining Yield Coverage Using Actual Production History § 1437.101 Actual production history. Actual production history (APH) is the unit's record of crop yield by crop year for the APH base period. The...

  15. Electron energy spectrum in cylindrical quantum dots and rods: approximation of separation of variables

    NASA Astrophysics Data System (ADS)

    Nedzinskas, R.; Karpus, V.; Čechavičius, B.; Kavaliauskas, J.; Valušis, G.

    2015-06-01

    A simple analytical method for electron energy spectrum calculations of cylindrical quantum dots (QDs) and quantum rods (QRs) is presented. The method is based on a replacement of an actual QD or QR hamiltonian with an approximate one, which allows for a separation of variables. Though this approach is known in the literature, it is essentially expanded in the present paper by taking into account a discontinuity of the effective mass, which is of importance in actual semiconductor heterostructures, e.g., InGaAs QDs or QRs embedded in GaAs matrix. Several examples of InGaAs QDs and QRs are considered—their energy spectrum calculations show that the suggested method yields reliable results both for the ground and excited states. The proposed analytical model is verified by numerical calculations, results of which coincide with an accuracy of ∼1 meV.

  16. The actual status of Astronomy in Moldova

    NASA Astrophysics Data System (ADS)

    Gaina, A.

    The astronomical research in the Republic of Moldova after Nicolae Donitch (Donici)(1874-1956(?)) were renewed in 1957, when a satellites observations station was open in Chisinau. Fotometric observations and rotations of first Soviet artificial satellites were investigated under a program SPIN put in action by the Academy of Sciences of former Socialist Countries. The works were conducted by Assoc. prof. Dr. V. Grigorevskij, which conducted also research in variable stars. Later, at the beginning of 60-th, an astronomical Observatory at the Chisinau State University named after Lenin (actually: the State University of Moldova), placed in Lozovo-Ciuciuleni villages was open, which were coordinated by Odessa State University (Prof. V.P. Tsesevich) and the Astrosovet of the USSR. Two main groups worked in this area: first conducted by V. Grigorevskij (till 1971) and second conducted by L.I. Shakun (till 1988), both graduated from Odessa State University. Besides this research areas another astronomical observations were made: Comets observations, astroclimate and atmospheric optics in collaboration with the Institute of the Atmospheric optics of the Siberian branch of the USSR (V. Chernobai, I. Nacu, C. Usov and A.F. Poiata). Comets observations were also made since 1988 by D. I. Gorodetskij which came to Chisinau from Alma-Ata and collaborated with Ukrainean astronomers conducted by K.I. Churyumov. Another part of space research was made at the State University of Tiraspol since the beggining of 70-th by a group of teaching staff of the Tiraspol State Pedagogical University: M.D. Polanuer, V.S. Sholokhov. No a collaboration between Moldovan astronomers and Transdniestrian ones actually exist due to War in Transdniestria in 1992. An important area of research concerned the Radiophysics of the Ionosphere, which was conducted in Beltsy at the Beltsy State Pedagogical Institute by a group of teaching staff of the University since the beginning of 70-th: N. D. Filip, E

  17. What Galvanic Vestibular Stimulation Actually Activates

    PubMed Central

    Curthoys, Ian S.; MacDougall, Hamish Gavin

    2012-01-01

    In a recent paper in Frontiers Cohen et al. (2012) asked “What does galvanic vestibular stimulation actually activate?” and concluded that galvanic vestibular stimulation (GVS) causes predominantly otolithic behavioral responses. In this Perspective paper we show that such a conclusion does not follow from the evidence. The evidence from neurophysiology is very clear: galvanic stimulation activates primary otolithic neurons as well as primary semicircular canal neurons (Kim and Curthoys, 2004). Irregular neurons are activated at lower currents. The answer to what behavior is activated depends on what is measured and how it is measured, including not just technical details, such as the frame rate of video, but the exact experimental context in which the measurement took place (visual fixation vs total darkness). Both canal and otolith dependent responses are activated by GVS. PMID:22833733

  18. MODIS Solar Diffuser: Modelled and Actual Performance

    NASA Technical Reports Server (NTRS)

    Waluschka, Eugene; Xiong, Xiao-Xiong; Esposito, Joe; Wang, Xin-Dong; Krebs, Carolyn (Technical Monitor)

    2001-01-01

    The Moderate Resolution Imaging Spectroradiometer (MODIS) instrument's solar diffuser is used in its radiometric calibration for the reflective solar bands (VIS, NTR, and SWIR) ranging from 0.41 to 2.1 micron. The sun illuminates the solar diffuser either directly or through a attenuation screen. The attenuation screen consists of a regular array of pin holes. The attenuated illumination pattern on the solar diffuser is not uniform, but consists of a multitude of pin-hole images of the sun. This non-uniform illumination produces small, but noticeable radiometric effects. A description of the computer model used to simulate the effects of the attenuation screen is given and the predictions of the model are compared with actual, on-orbit, calibration measurements.

  19. Alcohol based fuels for automotive engines

    SciTech Connect

    Menrad, H.K.

    1980-02-01

    The effects of methanol and/or ethanol additions on various properties of gasolines are discussed. Both advantages and disadvantages of such mixtures are set forth. The necessary changes in engine design to accommodate such fuel mixtures are described. Successful use of blended fuels in diesel engines is described. The current status of alcohol fuels in actual use is also reported. (BLM)

  20. Fuel Safety Activities in Korea

    SciTech Connect

    Auh, Geun-Sun; Shin, A.D.; Lee, J.S.; Woo, S.W.; Ryu, Y.H.; Kim, Jun-Hwan; Kim, S.K.; Jeong, Y.H.

    2007-07-01

    The current regulatory requirements for fuel performance were based on earlier test data of fresh or low burnup Zircaloy fuels of less than 40 GWD/MTU. Most countries have not changed the current regulatory requirements even if they are actively investigating the high burnup and new cladding alloy effects. Korea agrees with commonly accepted international consensus that although there are technical issues requiring resolutions, these issues do not constitute immediate safety concerns. The high burnup fuel reactor performance experiences of Korea do not show any major problems even if there have been some burnup related fuel failures which are described in the paper. KINS has recommended the industry to have lower fuel failure rates than 1-2 per 50,000 fuel rods. A research project of High Burnup Fuel Safety Tests and Evaluations has started in 2002 under a joint cooperation of KAERI/KNFC/KEPRI and KINS to obtain performance results of high burnup fuel and to develop evaluation technologies of high burnup fuel safety issues. From 1998, KINS has closely monitored and actively participated in international activities such as OECD/NEA CABRI Water Loop Program to reflect on regulatory requirements if needed. KINS will closely monitor the high burnup fuel performances of Korea to strength the regulatory activities if needed. The research activities in Korea including of LOCA and RIA being performed at KAERI with active supports of the industry are summarized in the paper. (authors)