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Sample records for reactor target rod

  1. Modeling the behavior of a light-water production reactor target rod

    SciTech Connect

    Sherwood, D.J.

    1992-03-01

    Pacific Northwest Laboratory has been conducting a series of in-reactor experiments in the Idaho National Engineering Laboratory (INEL) Advanced Test Reactor (ATR) to determine the amount of tritium released by permeation from a target rod under neutron irradiation. The model discussed in this report was developed from first principles to model the behavior of the first target rod irradiated in the ATR. The model can be used to determine predictive relationships for the amount of tritium that permeates through the target rod cladding during irradiation. The model consists of terms and equations for tritium production, gettering, partial pressure, and permeation, all of which are described in this report. The model addressed only the condition of steady state and features only a single adjustable parameter. The target rod design for producing tritium in a light-water reactor was tested first in the WC-1 in-reactor experiment. During irradiation, tritium is generated in the target rod within the ceramic lithium target material. The target rod has been engineered to limit the release of tritium to the reactor coolant during normal operations. The engineered features are a nickel-plated Zircaloy-4 getter and a barrier coating on the cladding surfaces. The ceramic target is wrapped with the getter material and the resulting ``pencils`` are inserted into the barrier coated cladding. These features of the rod are described in the report, along with the release of tritium from the ceramic target. The steady-state model could be useful for the design procedure of target rod components.

  2. Modeling the behavior of a light-water production reactor target rod

    SciTech Connect

    Sherwood, D.J.

    1992-03-01

    Pacific Northwest Laboratory has been conducting a series of in-reactor experiments in the Idaho National Engineering Laboratory (INEL) Advanced Test Reactor (ATR) to determine the amount of tritium released by permeation from a target rod under neutron irradiation. The model discussed in this report was developed from first principles to model the behavior of the first target rod irradiated in the ATR. The model can be used to determine predictive relationships for the amount of tritium that permeates through the target rod cladding during irradiation. The model consists of terms and equations for tritium production, gettering, partial pressure, and permeation, all of which are described in this report. The model addressed only the condition of steady state and features only a single adjustable parameter. The target rod design for producing tritium in a light-water reactor was tested first in the WC-1 in-reactor experiment. During irradiation, tritium is generated in the target rod within the ceramic lithium target material. The target rod has been engineered to limit the release of tritium to the reactor coolant during normal operations. The engineered features are a nickel-plated Zircaloy-4 getter and a barrier coating on the cladding surfaces. The ceramic target is wrapped with the getter material and the resulting pencils'' are inserted into the barrier coated cladding. These features of the rod are described in the report, along with the release of tritium from the ceramic target. The steady-state model could be useful for the design procedure of target rod components.

  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. 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.

  5. 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.

  6. Reactor control rod timing system

    SciTech Connect

    Wu, P.T.

    1982-02-09

    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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Rodded shutdown system for a nuclear reactor

    DOEpatents

    Golden, Martin P.; Govi, Aldo R.

    1978-01-01

    A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core.

  18. Nuclear reactor shutdown control rod assembly

    DOEpatents

    Bilibin, Konstantin

    1988-01-01

    A temperature responsive, self-actuated nuclear reactor shutdown control rod assembly 10. The upper end 18 of a lower drive line 17 fits within the lower end of an upper drive line 12. The lower end (not shown) of the lower drive line 17 is connected to a neutron absorber. During normal temperature conditions the lower drive line 17 is supported by detent means 22,26. When an overtemperature condition occurs thermal actuation means 34 urges ring 26 upwardly sufficiently to allow balls 22 to move radially outwardly thereby allowing lower drive line 17 to move downwardly toward the core of the nuclear reactor resulting in automatic reduction of the reactor powder.

  19. 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.

  20. 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.

  1. 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.

  2. CONTROL ROD FOR A NUCLEAR REACTOR AND METHOD OF PREPARATION

    DOEpatents

    Hausner, H.H.

    1958-12-30

    BS>An improved control rod is presented for a nuclear reactor. This control rod is comprised of a rare earth metal oxide or rare earth metal carbide such as gadolinium oxide or gadolinium carbide, uniformly distributed in a metal matrix having a low cross sectional area of absorption for thermal neutrons, such as aluminum, beryllium, and zirconium.

  3. Packed rod neutron shield for fast nuclear reactors

    DOEpatents

    Eck, John E.; Kasberg, Alvin H.

    1978-01-01

    A fast neutron nuclear reactor including a core and a plurality of vertically oriented neutron shield assemblies surrounding the core. Each assembly includes closely packed cylindrical rods within a polygonal metallic duct. The shield assemblies are less susceptible to thermal stresses and are less massive than solid shield assemblies, and are cooled by liquid coolant flow through interstices among the rods and duct.

  4. VARIABLE AREA CONTROL ROD FOR NUCLEAR REACTOR

    DOEpatents

    Huston, N.E.

    1960-05-01

    A control rod is described which permits continual variation of its absorbing strength uniformly along the length of the rod. The rod is fail safe and is fully inserted into the core but changes in its absorbing strength do not produce axial flux distortion. The control device comprises a sheet containing a material having a high thermal-neutron absorption cross section. A pair of shafts engage the sheet along the longitudinal axis of the shafts and gears associated with the shafts permit winding and unwinding of the sheet around the shafts.

  5. Pellet relocation testing results for four-foot-long tritium target rods

    SciTech Connect

    McKinnon, M.A.; Harding, N.E.

    1992-05-01

    This report discusses four-foot-long sections of a new production light-water reactor (NP-LWR) generic tritium target rod which were tested to determine if the length of the pellet pencils affects the amount of pellet material relocated during a burst and to characterize the burst. This testing was conducted as a follow-on study of cladding strength and pellet relocation behavior of short target rod specimens [11 cm (4-4 in.)]. The results of these tests could be used to support safety analyses of the effects of rod bursting and pellet relocation on the performance of a NP-LWR reactor core during a postulated loss-of-coolant accident (LOCA). All burst tests of the target rods were performed in air because air is more reactive than the air-steam or water environment that accompanies a LOCA.

  6. 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.

  7. 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.

  8. COAXIAL CONTROL ROD DRIVE MECHANISM FOR NEUTRONIC REACTORS

    DOEpatents

    Fox, R.J.; Oakes, L.C.

    1959-04-14

    A drive mechanism is presented for the control rod or a nuclear reactor. In this device the control rod is coupled to a drive shaft which extends coaxially through the rotor of an electric motor for relative rotation with respect thereto. A gear reduction mehanism is coupled between the rotor and the drive shaft to convert the rotary motion of the motor into linear motion of the shaft with a comparatively great reduction in speed, thereby providing relatively glow linear movement of the shaft and control rod for control purposes.

  9. ALLOY COMPOSITION FOR NEUTRONIC REACTOR CONTROL RODS

    DOEpatents

    Lustman, B.; Losco, E.F.; Snyder, H.J.; Eggleston, R.R.

    1963-01-22

    This invention relates to alloy compositons suitable as cortrol rod material consisting of, by weight, from 85% to 85% Ag, from 2% to 20% In, from up to 10% of Cd, from up to 5% Sn, and from up to 1.5% Al, the amount of each element employed being determined by the equation X + 2Y + 3Z + 3W + 4V = 1.4 and less, where X, Y, Z, W, and V represent the atom fractions of the elements Ag, Cd, In, Al and Sn. (AEC)

  10. Penetration of a copper rod into a sandy target

    NASA Astrophysics Data System (ADS)

    Kaminskii, M. V.; Kopytov, G. F.; Mogilev, V. A.; Travov, Yu. F.; Faikov, Yu. I.

    2010-05-01

    This paper presents the results of experimental and theoretical studies of high-velocity penetration of cylindrical copper rods into sand. The hydrodynamic Alekseevskii-Tate theory is modified to determine the penetration depth and wear velocity of the material of the rod penetrating into soil target in the plastic and hydrodynamic stages of penetration. The case where the target material is significantly less strong than the rod (impactor) material is considered.

  11. Fabrication of control rods for the High Flux Isotope Reactor

    SciTech Connect

    Sease, J.D.

    1998-03-01

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

  12. NEUTRONIC REACTOR CONTROL ROD AND METHOD OF FABRICATION

    DOEpatents

    Porembka, S.W. Jr.

    1961-06-27

    A reactor control rod formed from a compacted powder dispersion is patented. The rod consists of titanium sheathed with a cladding alloy. The cladding alloy contains 1.3% to 1.6% by weight of tin, 0.07% to 0.12% by weight of chromium, 0.04% to 0.08% by weight of nickel, 0.09% to 0.16% by weight of iron, carbon not exceeding 0.05%, less than 0.5% by weight of incidental impurities, and the balance zirconium.

  13. 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.

  14. 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).

  15. Fabrication of light water reactor tritium targets

    SciTech Connect

    Pilger, J.P.

    1991-11-01

    The mission of the Fabrication Development Task of the Tritium Target Development Project is: to produce a documented technology basis, including specifications and procedures for target rod fabrication; to demonstrate that light water tritium targets can be manufactured at a rate consistent with tritium production requirements; and to develop quality control methods to evaluate target rod components and assemblies, and establish correlations between evaluated characteristics and target rod performance. Many of the target rod components: cladding tubes, end caps, plenum springs, etc., have similar counterparts in LWR fuel rods. High production rate manufacture and inspection of these components has been adequately demonstrated by nuclear fuel rod manufacturers. This summary describes the more non-conventional manufacturing processes and inspection techniques developed to fabricate target rod components whose manufacturability at required production rates had not been previously demonstrated.

  16. 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

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

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

    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 fuel 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. 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.

  18. A two-step method for developing a control rod program for boiling water reactors

    SciTech Connect

    Taner, M.S.; Levine, S.H. ); Hsiao, M.Y. )

    1992-01-01

    This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in a computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift.

  19. Hybrid nuclear reactor grey rod to obtain required reactivity worth

    DOEpatents

    Miller, John V.; Carlson, William R.; Yarbrough, Michael B.

    1991-01-01

    Hybrid nuclear reactor grey rods are described, wherein geometric combinations of relatively weak neutron absorber materials such as stainless steel, zirconium or INCONEL, and relatively strong neutron absorber materials, such as hafnium, silver-indium cadmium and boron carbide, are used to obtain the reactivity worths required to reach zero boron change load follow. One embodiment includes a grey rod which has combinations of weak and strong neutron absorber pellets in a stainless steel cladding. The respective pellets can be of differing heights. A second embodiment includes a grey rod with a relatively thick stainless steel cladding receiving relatively strong neutron absorber pellets only. A third embodiment includes annular relatively weak netron absorber pellets with a smaller diameter pellet of relatively strong absorber material contained within the aperture of each relatively weak absorber pellet. The fourth embodiment includes pellets made of a homogeneous alloy of hafnium and a relatively weak absorber material, with the percentage of hafnium chosen to obtain the desired reactivity worth.

  20. NRC Targets University Reactors.

    ERIC Educational Resources Information Center

    Marshall, Eliot

    1984-01-01

    The Nuclear Regulatory Commission (NRC) wants universities to convert to low-grade fuel in their research reactions. Researchers claim the conversion, which will bring U.S. reactors in line with a policy the NRC is trying to impress on foreigners, could be financially and scientifically costly. Impact of the policy is considered. (JN)

  1. Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements.

    PubMed

    Kaiba, Tanja; Žerovnik, Gašper; Jazbec, Anže; Štancar, Žiga; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-10-01

    For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. PMID:26141293

  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. Control rod heterogeneity effects in liquid-metal fast breeder reactors: Method developments and experimental validation

    SciTech Connect

    Carta, M.; Granget, G.; Palmiotti, G.; Salvatores, M.; Soule, R.

    1988-11-01

    The control rod worth assessment in a large liquid-metal fast breeder reactor is strongly dependent on the actual arrangement of the absorber pins inside the control rod subassemblies. The so-called heterogeneity effects (i.e., the effects on the rod reactivity of the actual rod internal geometry versus homogenization of the absorber atoms over all the subassembly volume) have been evaluated, using explicit and variational methods to derive appropriate cross sections. An experimental program performed at the MASURCA facility has been used to validate these methods.

  4. Optimization of boiling water reactor control rod patterns using linear search

    SciTech Connect

    Kiguchi, T.; Doi, K.; Fikuzaki, T.; Frogner, B.; Lin, C.; Long, A.B.

    1984-10-01

    A computer program for searching the optimal control rod pattern has been developed. The program is able to find a control rod pattern where the resulting power distribution is optimal in the sense that it is the closest to the desired power distribution, and it satisfies all operational constraints. The search procedure consists of iterative uses of two steps: sensitivity analyses of local power and thermal margins using a three-dimensional reactor simulator for a simplified prediction model; linear search for the optimal control rod pattern with the simplified model. The optimal control rod pattern is found along the direction where the performance index gradient is the steepest. This program has been verified to find the optimal control rod pattern through simulations using operational data from the Oyster Creek Reactor.

  5. Materials and mechanical design analysis of boron carbide reactor safety rods

    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.

  6. 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.

  7. 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

  8. 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.

  9. Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

    SciTech Connect

    Pinto, Letícia Negrão; Gonnelli, Eduardo; Santos, Adimir dos

    2014-11-11

    Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of the control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.

  10. Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

    NASA Astrophysics Data System (ADS)

    Pinto, Letícia Negrão; Gonnelli, Eduardo; dos Santos, Adimir

    2014-11-01

    Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of the control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.

  11. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    SciTech Connect

    Richardson, K.D.

    1987-10-01

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580/sup 0/F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs.

  12. 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)

  13. Calculated concrete target damage by multiple rod impact and penetration

    SciTech Connect

    Pincosy, P A; Murphy, M J

    2006-12-29

    The effect of enhanced crater formation has been demonstrated experimentally when multiple and delayed shaped charge jets impact and penetrate concrete. The concept for enhancement utilizes a single follow-on jet at the centerline of holes produced by multiple precursor jets penetrating the surrounding the region. Calculations of the 3D crater enhancement phenomena have been conducted with multiple rods to simulate the steady state portion of the multiple jet penetration process. It is expected that this analysis methodology will be beneficial for optimization of the multiple jet crater enhancement application. We present calculated results using ALE3D where the model uses the standard Gruneisen equation of state combined with a rate dependent strength model including material damage parameters. This study focuses on the concrete material damage model as a representation of the portion of the target that would eventually be ejected creating a large bore-hole. The calculations are compared with the experimental evidence and limitations of the modeling approach are discussed.

  14. 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.

  15. 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

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. Performance evaluation of cigarette filter rods as a biofilm carrier in an anaerobic moving bed biofilm reactor.

    PubMed

    Sabzali, Ahmad; Nikaeen, Mahnaz; Bina, Bijan

    2012-01-01

    Biocarriers are an important component of anaerobic moving bed biofilm reactors (AMBBRs). In this study, the capability of cigarette filter rods (CFRs) as a biocarrier in an anaerobic moving bed biofilm reactor was evaluated. Two similar lab-scale anaerobic moving bed biofilm reactors were undertaken using Kaldnes-K3 plastic media and cigarette filter rods (wasted filters from tobacco factories) as biofilm attachment media for wastewater treatment. Organic substance and total posphours (TP) removal was investigated over 100 days. Synthetic wastewater was prepared with ordinary water and glucose as the main sources of carbon and energy, plus balanced macro- and micro-nutrients. Process performance was studied by increasing the organic loading rate (OLR) in the range of 1.6-4.5 kg COD/m3 x d. The COD average removal efficiency were 61.3% and 64.5% for AMBBR with cigarette filter rods (Reactor A) and AMBBR with Kaldnes plastic media (Reactor B), respectively. The results demonstrate that the performance of the AMBBR containing 0.25 litres of cigarette filters was comparable with a similar reactor containing 1.5 litres of Kaldnes plastic media. An average phosphorus removal of 67.7% and 72.9% was achieved by Reactors A and B, respectively.

  2. 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.

  3. 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.

  4. Aging mechanisms in the Westinghouse PWR (Pressurized Water Reactor) Control Rod Drive system

    SciTech Connect

    Gunther, W.; Sullivan, K.

    1991-01-01

    An aging assessment of the Westinghouse Pressurized Water Reactor (PWR) Control Rod System (CRD) has been completed as part of the US NRC's Nuclear Plant Aging Research, (NPAR) Program. This study examined the design, construction, maintenance, and operation of the system to determine its potential for degradation as the plant ages. Selected results from this study are presented in this paper. The operating experience data were evaluated to identify the predominant failure modes, causes, and effects. From our evaluation of the data, coupled with an assessment of the materials of construction and the operating environment, we conclude that the Westinghouse CRD system is subject to degradation which, if unchecked, could affect its safety function as a plant ages. Ways to detect and mitigate the effects of aging are included in this paper. The current maintenance for the control rod drive system at fifteen Westinghouse PWRs was obtained through a survey conducted in cooperation with EPRI and NUMARC. The results of the survey indicate that some plants have modified the system, replaced components, or expanded preventive maintenance. Several of these activities have effectively addressed the aging issue. 2 refs., 2 figs., 2 tabs.

  5. The impact of tungsten long rod penetrators into water filled targets

    SciTech Connect

    Wilson, L.T.; Dickinson, D.L.; Hertel, E.S. Jr.

    1998-02-01

    Twelve experiments were conducted to determine the effect of water filled targets on the penetration of tungsten long rods in terms of their residual mass and integrity. CTH hydrocode calculations were performed for each of the experiments to ensure that the erosion and breakup of the tungsten projectiles could be accurately reproduced. The CTH hydrocode predictions correlation well with the experimental results in most cases. Only 8% of the variance is unexplained. The slip interface between the rod and water was approximated in one of two ways: (1) using the CTH BLINT option in 2-D or (2) using a standard Eulerian mixed cells treatment. Results indicate that a 3-D BLINT algorithm is critical to predicting rod residual lengths. The authors were unable to reproduce rod fracture that occurred in every experiment where the water column exceeded 25 cm in length. The authors feel that this is due to a change in rod material properties during penetration, and continue to investigate the issue.

  6. Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

    DOE PAGESBeta

    Blau, P. J.; Qu, J.; Lu, R.

    2016-09-21

    One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes themore » basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.« less

  7. Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Blau, P. J.; Qu, J.; Lu, R.

    2016-09-01

    Fretting wear damage to fuel cladding from flow-induced vibrations can be a significant concern in the operation of light water nuclear reactors. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. The multi-stage model accounts for oxide layers and wear rate transitions. This paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

  8. Monte Carlo estimation of the dose and heating of cobalt adjuster rods irradiated in the CANDU 6 reactor core.

    PubMed

    Gugiu, Daniela; Dumitrache, Ion

    2005-01-01

    The present work is a part of a more complex project related to the replacement of the original stainless steel adjuster rods with cobalt assemblies in the CANDU 6 reactor core. The 60Co produced by 59Co irradiation could be used extensively in medicine and industry. The paper will mainly describe some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronic equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and heating of the irradiated cobalt rods, are performed using the Monte Carlo codes MCNP5 and MONTEBURNS 2.1. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters.

  9. 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.

  10. 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

  11. Laser-fusion targets for reactors

    DOEpatents

    Nuckolls, John H.; Thiessen, Albert R.

    1987-01-01

    A laser target comprising a thermonuclear fuel capsule composed of a centrally located quantity of fuel surrounded by at least one or more layers or shells of material for forming an atmosphere around the capsule by a low energy laser prepulse. The fuel may be formed as a solid core or hollow shell, and, under certain applications, a pusher-layer or shell is located intermediate the fuel and the atmosphere forming material. The fuel is ignited by symmetrical implosion via energy produced by a laser, or other energy sources such as an electron beam machine or ion beam machine, whereby thermonuclear burn of the fuel capsule creates energy for applications such as generation of electricity via a laser fusion reactor.

  12. External Attachment of Titanium Sheathed Thermocouples to Zirconium Nuclear Fuel Rods For The Loss-Of-Fluid-Test (LOFT) Reactor

    NASA Astrophysics Data System (ADS)

    Welty, Richard K.

    1980-10-01

    The Exxon Nuclear Company, Inc. acting as a Subcontractor to EG&G Idaho Inc.3 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). The design goals were to (1) reliably attach thermocouples to the zircaloy fuel rods, (2) achieve or exceed a life expectancy of 6,000 hours of reactor operation in a borated water environment of 316°C at 2260 psi, (3) provide and sustain repeatable physical and metallurgical properties in the instrumented rods subjected to transient temperatures up to 1538°C with blowdown, shock, loading, and fast quench. 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. Irradiation tests showed no degradation of thermocouples or weld structure. Fast thermal cycle and heater rod blowdown reflood tests were made to subject the weldments to high temperatures, high pressure steam, and fast water quench cycles. From the behavior of these tests, it was concluded that the attachment welds would survive a series of reactor safety tests.

  13. 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.

  14. Disposal Of Irradiated Cadmium Control Rods From The Plumbrook Reactor Facility

    SciTech Connect

    Posivak, E.J.; Berger, S.R.; Freitag, A.A.

    2008-07-01

    Innovative mixed waste disposition from NASA's Plum Brook Reactor Facility was accomplished without costly repackaging. Irradiated characteristic hardware with contact dose rates as high as 8 Sv/hr was packaged in a HDPE overpack and stored in a Secure Environmental Container during earlier decommissioning efforts, awaiting identification of a suitable pathway. WMG obtained regulatory concurrence that the existing overpack would serve as the macro-encapsulant per 40CFR268.45 Table 1.C. The overpack vent was disabled and the overpack was placed in a stainless steel liner to satisfy overburden slumping requirements. The liner was sealed and placed in shielded shoring for transport to the disposal site in a US DOT Type A cask. Disposition via this innovative method avoided cost, risk, and dose associated with repackaging the high dose irradiated characteristic hardware. In conclusion: WMG accomplished what others said could not be done. Large D and D contractors advised NASA that the cadmium control rods could only be shipped to the proposed Yucca mountain repository. NASA management challenged MOTA to find a more realistic alternative. NASA and MOTA turned to WMG to develop a methodology to disposition the 'hot and nasty' waste that presumably had no path forward. Although WMG lead a team that accomplished the 'impossible', the project could not have been completed with out the patient, supportive management by DOE-EM, NASA, and MOTA. (authors)

  15. Gas cooled fast reactor control rod drive mechanism deceleration unit. Test program

    SciTech Connect

    Wagner, T.H.

    1981-10-01

    This report presents the results of the airtesting portion of the proof-of-principle testing of a Control Rod Scram Deceleration Device developed for use in the Gas Cooled Fast Reactor (GCFR). The device utilizes a grooved flywheel to decelerate the translating assembly (T/A). Two cam followers on the translating assembly travel in the flywheel grooves and transfer the energy of the T/A to the flywheel. The grooves in the flywheel are straight for most of the flywheel length. Near the bottom of the T/A stroke the grooves are spiraled in a decreasing slope helix so that the cam followers accelerate the flywheel as they transfer the energy of the falling T/A. To expedite proof-of-principle testing, some of the materials used in the fabrication of certain test article components were not prototypic. With these exceptions the concept appears to be acceptable. The initial test of 300 scrams was completed with only one failure and the failure was that of a non-prototypic cam follower outer sleeve material.

  16. Dynamic inversion enables external magnets to concentrate ferromagnetic rods to a central target.

    PubMed

    Nacev, A; Weinberg, I N; Stepanov, P Y; Kupfer, S; Mair, L O; Urdaneta, M G; Shimoji, M; Fricke, S T; Shapiro, B

    2015-01-14

    The ability to use magnets external to the body to focus therapy to deep tissue targets has remained an elusive goal in magnetic drug targeting. Researchers have hitherto been able to manipulate magnetic nanotherapeutics in vivo with nearby magnets but have remained unable to focus these therapies to targets deep within the body using magnets external to the body. One of the factors that has made focusing of therapy to central targets between magnets challenging is Samuel Earnshaw's theorem as applied to Maxwell's equations. These mathematical formulations imply that external static magnets cannot create a stable potential energy well between them. We posited that fast magnetic pulses could act on ferromagnetic rods before they could realign with the magnetic field. Mathematically, this is equivalent to reversing the sign of the potential energy term in Earnshaw's theorem, thus enabling a quasi-static stable trap between magnets. With in vitro experiments, we demonstrated that quick, shaped magnetic pulses can be successfully used to create inward pointing magnetic forces that, on average, enable external magnets to concentrate ferromagnetic rods to a central location.

  17. Gold nano-rods as a targeting contrast agent for photoacoustic imaging

    NASA Astrophysics Data System (ADS)

    Agarwal, A.; Huang, S.-W.; Day, K. C.; O'Donnell, M.; Day, M.; Kotov, N.; Ashkenazi, S.

    2007-02-01

    We have studied the potential of gold nanorods to target cancer cells and provide contrast for photoacoustic imaging. The elongated "rod" shape of these nanoparticles provides a mechanism to tune their plasmon peak absorption wavelength. The absorption peak is shifted to longer wavelengths by increasing the aspect ratio of the rods. Particles 15 nm in diameter and 45 nm long were prepared using a seed mediated growth method. Their plasmon absorption peak was designed to be at 800 nm for increased penetration depth into biological tissue. They were conjugated with a specific antibody to target prostate cancer cells. We have applied photoacoustics to image a prostate cell culture targeted by conjugated gold particles. Images confirm the efficiency of conjugated particle binding to the targeted cell membranes. Photoacoustic detection of a single cell layer is demonstrated. To evaluate the applicability of the technique to clinical prostate cancer detection, we have imaged phantom objects mimicking a real tissue with small (2 mm size) inclusions of nanoparticle gel solution. Our photoacoustic imaging setup is based on a modified commercial ultrasonic scanner which makes it attractive for fast implementation in cancer diagnosis in clinical application. In addition, the setup allows for dual mode operation where a photoacoustic image is superimposed on a conventional B-mode ultrasound image. Dual mode operation is demonstrated by imaging a mouse with gold nanorod gel solution implanted in its hind limb.

  18. Reactor target from metal chromium for "pure" high-intensive artificial neutrino source

    NASA Astrophysics Data System (ADS)

    Gavrin, V. N.; Kozlova, Yu. P.; Veretenkin, E. P.; Logachev, A. V.; Logacheva, A. I.; Lednev, I. S.; Okunkova, A. A.

    2016-03-01

    The paper presents the first results of development of manufacturing technology of metallic chromium targets from highly enriched isotope 50Cr for irradiation in a high flux nuclear reactor to obtain a compact high intensity neutrino source with low content of radionuclide impurities and minimum losses of enriched isotope. The main technological stages are the hydrolysis of chromyl fluoride, the electrochemical reduction of metallic chromium, the hot isostatic pressing of chromium powder and the electrical discharge machining of chromium bars. The technological stages of hot isostatic pressing of chromium powder and of electrical discharge machining of Cr rods have been tested.

  19. Penetration of tungsten-alloy rods into composite ceramic targets: Experiments and 2-D simulations

    SciTech Connect

    Rosenberg, Z.; Dekel, E.; Hohler, V.; Stilp, A. J.; Weber, K.

    1998-07-10

    A series of terminal ballistics experiments, with scaled tungsten-alloy penetrators, was performed on composite targets consisting of ceramic tiles glued to thick steel backing plates. Tiles of silicon-carbide, aluminum nitride, titanium-dibroide and boron-carbide were 20-80 mm thick, and impact velocity was 1.7 km/s. 2-D numerical simulations, using the PISCES code, were performed in order to simulate these shots. It is shown that a simplified version of the Johnson-Holmquist failure model can account for the penetration depths of the rods but is not enough to capture the effect of lateral release waves on these penetrations.

  20. CFD simulation of an unbaffled stirred tank reactor driven by a magnetic rod: assessment of turbulence models.

    PubMed

    Li, Jiajia; Deng, Baoqing; Zhang, Bing; Shen, Xiuzhong; Kim, Chang Nyung

    2015-01-01

    A simulation of an unbaffled stirred tank reactor driven by a magnetic stirring rod was carried out in a moving reference frame. The free surface of unbaffled stirred tank was captured by Euler-Euler model coupled with the volume of fluid (VOF) method. The re-normalization group (RNG) k-ɛ model, large eddy simulation (LES) model and detached eddy simulation (DES) model were evaluated for simulating the flow field in the stirred tank. All turbulence models can reproduce the tangential velocity in an unbaffled stirred tank with a rotational speed of 150 rpm, 250 rpm and 400 rpm, respectively. Radial velocity is underpredicted by the three models. LES model and RNG k-ɛ model predict the better tangential velocity and axial velocity, respectively. RNG k-ɛ model is recommended for the simulation of the flow in an unbaffled stirred tank with magnetic rod due to its computational effort.

  1. 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.

  2. cAMP controls rod photoreceptor sensitivity via multiple targets in the phototransduction cascade

    PubMed Central

    Astakhova, Luba A.; Samoiliuk, Evgeniia V.; Govardovskii, Victor I.

    2012-01-01

    In early studies, both cyclic AMP (cAMP) and cGMP were considered as potential secondary messengers regulating the conductivity of the vertebrate photoreceptor plasma membrane. Later discovery of the cGMP specificity of cyclic nucleotide–gated channels has shifted attention to cGMP as the only secondary messenger in the phototransduction cascade, and cAMP is not considered in modern schemes of phototransduction. Here, we report evidence that cAMP may also be involved in regulation of the phototransduction cascade. Using a suction pipette technique, we recorded light responses of isolated solitary rods from the frog retina in normal solution and in the medium containing 2 µM of adenylate cyclase activator forskolin. Under forskolin action, flash sensitivity rose more than twofold because of a retarded photoresponse turn-off. The same concentration of forskolin lead to a 2.5-fold increase in the rod outer segment cAMP, which is close to earlier reported natural day/night cAMP variations. Detailed analysis of cAMP action on the phototransduction cascade suggests that several targets are affected by cAMP increase: (a) basal dark phosphodiesterase (PDE) activity decreases; (b) at the same intensity of light background, steady background-induced PDE activity increases; (c) at light backgrounds, guanylate cyclase activity at a given fraction of open channels is reduced; and (d) the magnitude of the Ca2+ exchanger current rises 1.6-fold, which would correspond to a 1.6-fold elevation of [Ca2+]in. Analysis by a complete model of rod phototransduction suggests that an increase of [Ca2+]in might also explain effects (b) and (c). The mechanism(s) by which cAMP could regulate [Ca2+]in and PDE basal activity is unclear. We suggest that these regulations may have adaptive significance and improve the performance of the visual system when it switches between day and night light conditions. PMID:23008435

  3. cAMP controls rod photoreceptor sensitivity via multiple targets in the phototransduction cascade.

    PubMed

    Astakhova, Luba A; Samoiliuk, Evgeniia V; Govardovskii, Victor I; Firsov, Michael L

    2012-10-01

    In early studies, both cyclic AMP (cAMP) and cGMP were considered as potential secondary messengers regulating the conductivity of the vertebrate photoreceptor plasma membrane. Later discovery of the cGMP specificity of cyclic nucleotide-gated channels has shifted attention to cGMP as the only secondary messenger in the phototransduction cascade, and cAMP is not considered in modern schemes of phototransduction. Here, we report evidence that cAMP may also be involved in regulation of the phototransduction cascade. Using a suction pipette technique, we recorded light responses of isolated solitary rods from the frog retina in normal solution and in the medium containing 2 µM of adenylate cyclase activator forskolin. Under forskolin action, flash sensitivity rose more than twofold because of a retarded photoresponse turn-off. The same concentration of forskolin lead to a 2.5-fold increase in the rod outer segment cAMP, which is close to earlier reported natural day/night cAMP variations. Detailed analysis of cAMP action on the phototransduction cascade suggests that several targets are affected by cAMP increase: (a) basal dark phosphodiesterase (PDE) activity decreases; (b) at the same intensity of light background, steady background-induced PDE activity increases; (c) at light backgrounds, guanylate cyclase activity at a given fraction of open channels is reduced; and (d) the magnitude of the Ca(2+) exchanger current rises 1.6-fold, which would correspond to a 1.6-fold elevation of [Ca(2+)](in). Analysis by a complete model of rod phototransduction suggests that an increase of [Ca(2+)](in) might also explain effects (b) and (c). The mechanism(s) by which cAMP could regulate [Ca(2+)](in) and PDE basal activity is unclear. We suggest that these regulations may have adaptive significance and improve the performance of the visual system when it switches between day and night light conditions. PMID:23008435

  4. Application of cigarette filter rods as biofilm carrier in an integrated fixed-film activated sludge reactor.

    PubMed

    Sabzali, Ahmad; Nikaeen, Mahnaz; Bina, Bijan

    2013-01-01

    Bio-carriers are an important component of integrated fixed-film activated sludge (IFAS) processes. In this study, the capability of cigarette filter rods (CFRs) as a bio-carrier in IFAS processes was evaluated. Two similar laboratory-scale IFAS systems were operated over a 4-month period using Kaldnes-K3 and CFRs as IFAS media. The process performance was studied by using chemical oxygen demand (COD). The organic loading rate was in the range 0.5-2.8 kgCOD/(m(3)·d). The COD average removal efficiencies were 89.3 and 93.9% for Kaldnes-K3 (reactor A) and cigarette filters (reactor B), respectively. The results demonstrate that the performance of the IFAS reactor containing CFRs was comparable to the reactor using Kaldnes. The CFRs, which have a high porous surface area and entrapment ability for microbial cells, could be successfully used in biofilm reactors as a bio-carrier.

  5. Scrapie prion liposomes and rods exhibit target sizes of 55,000 Da

    SciTech Connect

    Bellinger-Kawahara, C.G.; Kempner, E.; Groth, D.; Gabizon, R.; Prusiner, S.B.

    1988-06-01

    Scrapie is a degenerative neurologic disease in sheep and goats which can be experimentally transmitted to laboratory rodents. Considerable evidence suggests that the scrapie agent is composed largely, if not entirely, of an abnormal isoform of the prion protein (PrPSc). Inactivation of scrapie prions by ionizing radiation exhibited single-hit kinetics and gave a target size of 55,000 +/- 9000 mol wt. The inactivation profile was independent of the form of the prion. Scrapie agent infectivity in brain homogenates, microsomal fractions, detergent-extracted microsomes, purified amyloid rods, and liposomes exhibited the same inactivation profile. Our data are consistent with the hypothesis that the infectious particle causing scrapie contains approximately 2 PrPSc molecules.

  6. 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.

  7. Thermal hydraulic performance of naturally aspirated control rod housing assemblies

    SciTech Connect

    Geiger, G.T.; Randolph, H.W.; Paik, I.K.; Foti, D.J.

    1992-08-01

    Savannah River Site reactors are comprised of heat generating fuel/target assemblies, control rods which regulate reactor power, and heavy water which acts as the coolant and as a moderator. The fuel/target assemblies are cooled by the downflow of heavy water while the control rods are cooled via upflow. Five control rods are grouped with two safety rods in seven-channel assemblies called septifoils. Under normal operating conditions, the reactor power level, radial shape flux and axial power flux are regulated by the positioning of the control rods. The control rods are solid rods of a lithium-aluminum alloy with an thin aluminum outer sheath. Lithium is a good absorber of neutrons and, thus control rod temperatures rise with reactor power. At conditions of sufficiently high reactor power and degraded coolant flow, the control rods could heat sufficiently to cause a metallurigical failure of the sheath leading to molten material coming in contact with water and the possibility of a steam explosion. An accident has been postulated as part of the analysis involving the safety upgrade of Savannah River Site reactors in which the housing is not seated on the pin. Coolant from the upflow pin would not be directed into the housing but, into the moderator space surrounding the housing. Only naturally aspirated cooling due to buoyancy effects would be available to cool the control rods and the coolant mass flow rate would drop significantly from its nominal value. In this study, the mechanisms and limits of cooling heated rods housed in an unseated septifoil are addressed. Experiments were conducted on a shortened, prototypic housing with electrically heated rods to gain an understanding of the phenomena governing the cooling in such a case and develop data which can be used to evaluate predictive models. These experiments are described, their results discussed, and the predictions of current models is presented.

  8. Thermal hydraulic performance of naturally aspirated control rod housing assemblies

    SciTech Connect

    Geiger, G.T.; Randolph, H.W.; Paik, I.K. ); Foti, D.J. )

    1992-01-01

    Savannah River Site reactors are comprised of heat generating fuel/target assemblies, control rods which regulate reactor power, and heavy water which acts as the coolant and as a moderator. The fuel/target assemblies are cooled by the downflow of heavy water while the control rods are cooled via upflow. Five control rods are grouped with two safety rods in seven-channel assemblies called septifoils. Under normal operating conditions, the reactor power level, radial shape flux and axial power flux are regulated by the positioning of the control rods. The control rods are solid rods of a lithium-aluminum alloy with an thin aluminum outer sheath. Lithium is a good absorber of neutrons and, thus control rod temperatures rise with reactor power. At conditions of sufficiently high reactor power and degraded coolant flow, the control rods could heat sufficiently to cause a metallurigical failure of the sheath leading to molten material coming in contact with water and the possibility of a steam explosion. An accident has been postulated as part of the analysis involving the safety upgrade of Savannah River Site reactors in which the housing is not seated on the pin. Coolant from the upflow pin would not be directed into the housing but, into the moderator space surrounding the housing. Only naturally aspirated cooling due to buoyancy effects would be available to cool the control rods and the coolant mass flow rate would drop significantly from its nominal value. In this study, the mechanisms and limits of cooling heated rods housed in an unseated septifoil are addressed. Experiments were conducted on a shortened, prototypic housing with electrically heated rods to gain an understanding of the phenomena governing the cooling in such a case and develop data which can be used to evaluate predictive models. These experiments are described, their results discussed, and the predictions of current models is presented.

  9. Control rod drive

    DOEpatents

    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.

  10. Analytical models for the penetration of semi-infinite targets by rigid, deformable and erosive long rods

    NASA Astrophysics Data System (ADS)

    Wen, He-Ming; Lan, Bin

    2010-08-01

    A theoretical study is presented herein on the penetration of a semi-infinite target by a spherical-headed long rod for Y p > S, where Y p is the penetrator strength and S is the static target resistance. For Y p > S, depending upon initial impact velocity, there exist three types of penetration, namely, penetration by a rigid long rod, penetration by a deforming non-erosive long rod and penetration by an erosive long rod. If the impact velocity of the penetrator is higher than the hydrodynamic velocity ( V H), it will penetrate the target in an erosive mode; if the impact velocity lies between the hydrodynamic velocity ( V H) and the rigid body velocity ( V R), it will penetrate the target in a deformable mode; if the impact velocity is less than the rigid body velocity ( V R), it will penetrate the target in a rigid mode. The critical conditions for the transition among these three penetration modes are proposed. It is demonstrated that the present model predictions correlate well with the experimental observations in terms of depth of penetration (DOP) and the critical transition conditions.

  11. Generation of a Nanobody Targeting the Paraflagellar Rod Protein of Trypanosomes

    PubMed Central

    Obishakin, Emmanuel; Stijlemans, Benoit; Santi-Rocca, Julien; Vandenberghe, Isabel; Devreese, Bart; Muldermans, Serge; Bastin, Philippe; Magez, Stefan

    2014-01-01

    Trypanosomes are protozoan parasites that cause diseases in humans and livestock for which no vaccines are available. Disease eradication requires sensitive diagnostic tools and efficient treatment strategies. Immunodiagnostics based on antigen detection are preferable to antibody detection because the latter cannot differentiate between active infection and cure. Classical monoclonal antibodies are inaccessible to cryptic epitopes (based on their size-150 kDa), costly to produce and require cold chain maintenance, a condition that is difficult to achieve in trypanosomiasis endemic regions, which are mostly rural. Nanobodies are recombinant, heat-stable, small-sized (15 kDa), antigen-specific, single-domain, variable fragments derived from heavy chain-only antibodies in camelids. Because of numerous advantages over classical antibodies, we investigated the use of nanobodies for the targeting of trypanosome-specific antigens and diagnostic potential. An alpaca was immunized using lysates of Trypanosoma evansi. Using phage display and bio-panning techniques, a cross-reactive nanobody (Nb392) targeting all trypanosome species and isolates tested was selected. Imunoblotting, immunofluorescence microscopy, immunoprecipitation and mass spectrometry assays were combined to identify the target recognized. Nb392 targets paraflagellar rod protein (PFR1) of T. evansi, T. brucei, T. congolense and T. vivax. Two different RNAi mutants with defective PFR assembly (PFR2RNAi and KIF9BRNAi) were used to confirm its specificity. In conclusion, using a complex protein mixture for alpaca immunization, we generated a highly specific nanobody (Nb392) that targets a conserved trypanosome protein, i.e., PFR1 in the flagella of trypanosomes. Nb392 is an excellent marker for the PFR and can be useful in the diagnosis of trypanosomiasis. In addition, as demonstrated, Nb392 can be a useful research or PFR protein isolation tool. PMID:25551637

  12. The necessity of nuclear reactors for targeted radionuclide therapies.

    PubMed

    Krijger, Gerard C; Ponsard, Bernard; Harfensteller, Mark; Wolterbeek, Hubert T; Nijsen, Johannes W F

    2013-07-01

    Nuclear medicine has been contributing towards personalized therapies. Nuclear reactors are required for the working horses of both diagnosis and treatment, i.e., Tc-99m and I-131. In fact, reactors will remain necessary to fulfill the demand for a variety of radionuclides and are essential in the expanding field of targeted radionuclide therapies for cancer. However, the main reactors involved in the global supply are ageing and expected to shut down before 2025. Therefore, the fields of (nuclear) medicine, nuclear industry and politics share a global responsibility, faced with the task to secure future access to suitable nuclear reactors. At the same time, alternative production routes should be industrialized. For this, a coordinating entity should be put into place.

  13. Computer program for automatic generation of BWR control rod patterns

    SciTech Connect

    Taner, M.S.; Levine, S.H.; Hsia, M.Y. )

    1990-01-01

    A computer program named OCTOPUS has been developed to automatically determine a control rod pattern that approximates some desired target power distribution as closely as possible without violating any thermal safety or reactor criticality constraints. The program OCTOPUS performs a semi-optimization task based on the method of approximation programming (MAP) to develop control rod patterns. The SIMULATE-E code is used to determine the nucleonic characteristics of the reactor core state.

  14. 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

  15. A rule-based expert system for automatic control rod pattern generation for boiling water reactors

    SciTech Connect

    Lin, L.S.; Lin, C. )

    1991-07-01

    This paper reports on an expert system for generating control rod patterns that has been developed. The knowledge is transformed into IF-THEN rules. The inference engine uses the Rete pattern matching algorithm to match facts, and rule premises and conflict resolution strategies to make the system function intelligently. A forward-chaining mechanism is adopted in the inference engine. The system is implemented in the Common Lisp programming language. The three-dimensional core simulation model performs the core status and burnup calculations. The system is successfully demonstrated by generating control rod programming for the 2894-MW (thermal) Kuosheng nuclear power plant in Taiwan. The computing time is tremendously reduced compared to programs using mathematical methods.

  16. Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Solonin, V. I.; Perevezentsev, V. V.

    2012-05-01

    Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.

  17. Study of material changes of SINQ target rods after long-term exposure by neutron radiography methods.

    PubMed

    Lehmann, E E; Vontobel, P; Estermann, M

    2004-10-01

    This paper describes the results of non-destructive investigations by indirect neutron radiography methods obtained at the facility NEUTRA [Nondestruct. Testing Eval. 16 (2000b) 203], spallation neutron source SINQ [Operating experience and development projects at SINQ, PSI Report 98-04, ISSN 1019-0643]. Target rods from the second SINQ metal target were removed after 6 Ah of proton beam exposure and studied under well-shielded conditions. No real damage was found at one of the 11 observed rods and one tube. However, hydrogen accumulation could be identified inside the zircaloy rods and the steel rods as well. Whereas the hydrogen has a homogenous distribution in Zr (with the peak value near the centre of the applied beam), the steel samples show clusters of hydrogen near the edge of the Zr cladding. Lead (in steel cladding) was found modified by accumulations of spallation products, mainly mercury. In the radiography images, a depression of the neutron field was observed due to the absorption by mercury. The applied method with Dy and In as neutron converters and imaging plates [Nucl. Instrum. Methods 377 (1996) 119] as secondary detectors seems to be optimal for such kind of investigations, especially when quantitative considerations have to be made. PMID:15246406

  18. An effective loading method of americium targets in fast reactors

    SciTech Connect

    Ohki, Shigeo; Sato, Isamu; Mizuno, Tomoyasu; Hayashi, Hideyuki; Tanaka, Kenya

    2007-07-01

    Recently, the development of target fuel with high americium (Am) content has been launched for the reduction of the overall fuel fabrication cost of the minor actinide (MA) recycling. In the framework of the development, this study proposes an effective loading method of Am targets in fast reactors. As a result of parametric survey calculations, we have found the ring-shaped target loading pattern between inner and outer core regions. This loading method is satisfactory both in core characteristics and in MA transmutation property. It should be noted that the Am targets can contribute to the suppression of the core power distribution change due to burnup. The major drawback of Am target is the production of helium gas. A target design modification by increasing the cladding thickness is found to be the most feasible measure to cope with the helium production. (authors)

  19. 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.

  20. 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.

  1. 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.

  2. Rod sequence advisor

    SciTech Connect

    Wood, R.M. ); Lu, Yi ); Furia, R.V.; Thompson, R.J. ); Lin, Ching-lu )

    1992-01-01

    During startup and power shaping maneuvers of boiling water reactors (BWR's), control rods are sequentially withdrawn from the reactor core. The withdrawal sequences determine the overall reactor power and the local core power density and are based on the knowledge of station engineers. It is important that the control rods are withdrawn in such a manner that the local power level does not become excessive while the desired reactor power is generated. Rules that constrain the relative positions of control rod groups have been developed to do this. While these rules are relatively simple, applying them to all possible movements of the 17 control rod groups in a typical BWR is complex and time consuming. SMARTRODS, is a rule based pilot expert system, was developed in LISP for the determination of the rod sequences.

  3. DESCRIPTION OF THE TRITIUM-PRODUCING BURNABLE ABSORBER ROD FOR THE COMMERCIAL LIGHT WATER REACTOR TTQP-1-015 Rev 19

    SciTech Connect

    Burns, Kimberly A.; Love, Edward F.; Thornhill, Cheryl K.

    2012-02-01

    Tritium-producing burnable absorber rods (TPBARs) used in the U.S. Department of Energy’s Tritium Readiness Program are designed to produce tritium when placed in a Westinghouse or Framatome 17x17 fuel assembly and irradiated in a pressurized water reactor (PWR). This document provides an unclassified description of the current design baseline for the TPBARs. This design baseline is currently valid only for Watts Bar reactor production cores. A description of the Lead Use TPBARs will not be covered in the text of the document, but the applicable drawings, specifications and test plan will be included in the appropriate appendices.

  4. Experimental investigations on decay heat removal in advanced nuclear reactors using single heater rod test facility: Air alone in the annular gap

    SciTech Connect

    Bopche, Santosh B.; Sridharan, Arunkumar

    2010-11-15

    During a loss of coolant accident in nuclear reactors, radiation heat transfer accounts for a significant amount of the total heat transfer in the fuel bundle. In case of heavy water moderator nuclear reactors, the decay heat of a fuel bundle enclosed in the pressure tube and outer concentric calandria tube can be transferred to the moderator. Radiation heat transfer plays a significant role in removal of decay heat from the fuel rods to the moderator, which is available outside the calandria tube. A single heater rod test facility is designed and fabricated as a part of preliminary investigations. The objective is to anticipate the capability of moderator to remove decay heat, from the reactor core, generated after shut down. The present paper focuses mainly on the role of moderator in removal of decay heat, for situation with air alone in the annular gap of pressure tube and calandria tube. It is seen that the naturally aspirated air is capable of removing the heat generated in the system compared to the standstill air or stagnant water situations. It is also seen that the flowing moderator is capable of removing a greater fraction of heat generated by the heater rod compared to a stagnant pool of boiling moderator. (author)

  5. 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)

  6. Results from studies of surface deposits on the claddings of fuel rods used in RBMK-1000 reactors

    NASA Astrophysics Data System (ADS)

    Smirnova, I. M.; Markov, D. V.

    2010-07-01

    The results of studies on analyzing the element composition of deposits on the cladding surfaces of fuel rods used in a fuel assembly at the Leningrad nuclear power station are presented. The distribution of elements in deposits over the fuel rod height is analyzed, and the zones of their concentration are revealed. It is shown that deposits of copper penetrating into cracks in the surface layer of zirconium oxide introduce an essential contribution in the development of nodular corrosion of fuel rod claddings.

  7. Reactor Physics Characterization of Transmutation Targeting Options in a Sodium Fast Reactor

    SciTech Connect

    Samuel E. Bays

    2007-04-01

    In sodium fast reactor designs, the fuel related inherent negative reactivity feedback is accomplished mainly through parasitic capture in U-238. However for an efficient minor actinide burning system, it is desirable to reduce or eliminate U-238 entirely to suppress further transuranic actinide generation. Consequently, reactivity feedback is accomplished by enhancing axial neutron streaming during a loss of coolant void situation. This is done by flattening “pancake” the active core geometry. Flattening the reactor also increases axial leakage which removes neutrons that could otherwise be used to destroy minor actinides. Therefore, it is important to tailor the neutron spectrum in the core for optimized feedback and minor actinide destruction simultaneously by using minor actinide and fission product targets.

  8. 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. ...

  9. 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.

  10. Understanding the Atomic-Level Chemistry and Structure of Oxide Deposits on Fuel Rods in Light Water Nuclear Reactors Using First Principles Methods

    NASA Astrophysics Data System (ADS)

    Rak, Zs.; O'Brien, C. J.; Brenner, D. W.; Andersson, D. A.; Stanek, C. R.

    2016-09-01

    The results of recent studies are discussed in which first principles calculations at the atomic level have been used to expand the thermodynamic database for science-based predictive modeling of the chemistry, composition and structure of unwanted oxides that deposit on the fuel rods in pressurized light water nuclear reactors. Issues discussed include the origin of the particles that make up deposits, the structure and properties of the deposits, and the forms by which boron uptake into the deposits can occur. These first principles approaches have implications for other research areas, such as hydrothermal synthesis and the stability and corrosion resistance of other materials under other extreme conditions.

  11. Flagellar Morphogenesis: Protein Targeting and Assembly in the Paraflagellar Rod of Trypanosomes

    PubMed Central

    Bastin, Philippe; MacRae, Thomas H.; Francis, Susan B.; Matthews, Keith R.; Gull, Keith

    1999-01-01

    The paraflagellar rod (PFR) of the African trypanosome Trypanosoma brucei represents an excellent model to study flagellum assembly. The PFR is an intraflagellar structure present alongside the axoneme and is composed of two major proteins, PFRA and PFRC. By inducible expression of a functional epitope-tagged PFRA protein, we have been able to monitor PFR assembly in vivo. As T. brucei cells progress through their cell cycle, they possess both an old and a new flagellum. The induction of expression of tagged PFRA in trypanosomes growing a new flagellum provided an excellent marker of newly synthesized subunits. This procedure showed two different sites of addition: a major, polar site at the distal tip of the flagellum and a minor, nonpolar site along the length of the partially assembled PFR. Moreover, we have observed turnover of epitope-tagged PFRA in old flagella that takes place throughout the length of the PFR structure. Expression of truncated PFRA mutant proteins identified a sequence necessary for flagellum localization by import or binding. This sequence was not sufficient to confer full flagellum localization to a green fluorescent protein reporter. A second sequence, necessary for the addition of PFRA protein to the distal tip, was also identified. In the absence of this sequence, the mutant PFRA proteins were localized both in the cytosol and in the flagellum where they could still be added along the length of the PFR. This seven-amino-acid sequence is conserved in all PFRA and PFRC proteins and shows homology to a sequence in the flagellar dynein heavy chain of Chlamydomonas reinhardtii. PMID:10567544

  12. 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)

  13. Rod guide

    SciTech Connect

    Sable, D.E.

    1988-11-29

    This patent describes a rod guide assembly for a sucker rod longitudinally reciprocably movable in a well flow conductor comprising: a pair of longitudinally spaced upper and lower stops rigidly secured to a sucker rod; and a guide body movably mounted on the rod between the stops. The stops being spaced from each other a distance slightly greater than the length of the guide body, the upper stop engaging the guide body to move the guide body downwardly with the rod after an initial short downward movement of the rod after initiation of each downward movement of the rod and the lower stop engaging the guide body to move the second guide body upwardly with the rod after initial short upward movement of the rod after initiation of each upward movement of the rod during the longitudinal reciprocatory movement of the rod in a well flow conductor.

  14. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  15. Synergistic Targeting of Cell Membrane, Cytoplasm and Nucleus of Cancer Cells using Rod-Shaped Nanoparticles

    PubMed Central

    Barua, Sutapa; Mitragotri, Samir

    2014-01-01

    Design of carriers for effective delivery and targeting of drugs to cellular and sub-cellular compartments is an unmet need in medicine. Here, we report pure drug nanoparticles comprising camptothecin (CPT), trastuzumab (TTZ) and doxorubicin (DOX) to enable cell-specific interactions, subcellular accumulation and growth inhibition of breast cancer cells. CPT is formulated in the form of nanorods which are coated with TTZ. DOX is encapsulated in the TTZ corona around the CPT nanoparticle. Our results show that TTZ/DOX-coated CPT nanorods exhibit cell-specific internalization in BT-474 breast cancer cells, after which TTZ is recycled to the plasma membrane leaving CPT nanorods in the perinuclear region and delivering DOX into the nucleus of the cells. The effects of CPT-TTZ-DOX nanoparticles on growth inhibition are synergistic (combination index = 0.17±0.03) showing 10-10,000 fold lower inhibitory concentrations (IC50) compared to those of individual drugs. The design of antibody-targeted pure drug nanoparticles offers a promising design strategy to facilitate intracellular delivery and therapeutic efficiency of anticancer drugs. PMID:24053162

  16. Detection of Target Biomolecules by Magnetic Reporting Using Rod-Like Nanosensors

    NASA Astrophysics Data System (ADS)

    Guertin, R. P.; Goldberg, E.; Harrah, T. P.; Sonkusale, S.; Park, K.; Sun, S.; Oh, J. I.; Naughton, M.

    2008-03-01

    We describe the ongoing development of a device to assay a variety of cellular, viral and molecular targets by measuring the increase of the Brownian relaxation time, τ, in solution of magnetically-tagged nanoscale detectors. The shift shows as a frequency reduction of the peak of the complex magnetic susceptibility, χ(φ)''. Measurements of χ(φ)'' with 12 nm monodisperse nanoparticles of CoFe2O4 coated with polyethelyne glycol reveal spectra with the narrowest lines yet reported. Thin avidin coating of these particles reveals small shifts in χ(φ)''. Bacteriophage T4 tail fibers, engineered to specific lengths (30-150 nm), were employed as the platform for magnetic nanoparticle attachment and at the other end for an inserted target peptide epitope. Attachment of the nanoparticles to bacteriophage T4 tail fibers was successful, though no detectable shifts in χ(φ)'' were detected due to weak attachment. The advantages associated with non-spherical geometry detectors will be discussed, as will preliminary measurements with rare earth oxide magnetic nanoparticles. Progress on miniaturization and low power requirements of the electronic detection system will be reported. Supported by NERCE/BEID (NIAID).

  17. Synergistic targeting of cell membrane, cytoplasm, and nucleus of cancer cells using rod-shaped nanoparticles.

    PubMed

    Barua, Sutapa; Mitragotri, Samir

    2013-11-26

    Design of carriers for effective delivery and targeting of drugs to cellular and subcellular compartments is an unmet need in medicine. Here, we report pure drug nanoparticles comprising camptothecin (CPT), trastuzumab (TTZ), and doxorubicin (DOX) to enable cell-specific interactions, subcellular accumulation, and growth inhibition of breast cancer cells. CPT is formulated in the form of nanorods which are coated with TTZ. DOX is encapsulated in the TTZ corona around the CPT nanoparticle. Our results show that TTZ/DOX-coated CPT nanorods exhibit cell-specific internalization in BT-474 breast cancer cells, after which TTZ is recycled to the plasma membrane, leaving CPT nanorods in the perinuclear region and delivering DOX into the nucleus of the cells. The effects of CPT-TTZ-DOX nanoparticles on growth inhibition are synergistic (combination index = 0.17 ± 0.03) showing 10-10 000-fold lower inhibitory concentrations (IC50) compared to those of individual drugs. The design of antibody-targeted pure drug nanoparticles offers a promising design strategy to facilitate intracellular delivery and therapeutic efficiency of anticancer drugs.

  18. 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)

  19. Penetration of rod projectiles in semi-infinite targets : a validation test for Eulerian X-FEM in ALEGRA.

    SciTech Connect

    Park, Byoung Yoon; Leavy, Richard Brian; Niederhaus, John Henry J.

    2013-03-01

    The finite-element shock hydrodynamics code ALEGRA has recently been upgraded to include an X-FEM implementation in 2D for simulating impact, sliding, and release between materials in the Eulerian frame. For validation testing purposes, the problem of long-rod penetration in semi-infinite targets is considered in this report, at velocities of 500 to 3000 m/s. We describe testing simulations done using ALEGRA with and without the X-FEM capability, in order to verify its adequacy by showing X-FEM recovers the good results found with the standard ALEGRA formulation. The X-FEM results for depth of penetration differ from previously measured experimental data by less than 2%, and from the standard formulation results by less than 1%. They converge monotonically under mesh refinement at first order. Sensitivities to domain size and rear boundary condition are investigated and shown to be small. Aside from some simulation stability issues, X-FEM is found to produce good results for this classical impact and penetration problem.

  20. SNAREs Interact with Retinal Degeneration Slow and Rod Outer Segment Membrane Protein-1 during Conventional and Unconventional Outer Segment Targeting

    PubMed Central

    Zulliger, Rahel; Conley, Shannon M.; Mwoyosvi, Maggie L.; Stuck, Michael W.; Azadi, Seifollah; Naash, Muna I.

    2015-01-01

    Mutations in the photoreceptor protein peripherin-2 (also known as RDS) cause severe retinal degeneration. RDS and its homolog ROM-1 (rod outer segment protein 1) are synthesized in the inner segment and then trafficked into the outer segment where they function in tetramers and covalently linked larger complexes. Our goal is to identify binding partners of RDS and ROM-1 that may be involved in their biosynthetic pathway or in their function in the photoreceptor outer segment (OS). Here we utilize several methods including mass spectrometry after affinity purification, in vitro co-expression followed by pull-down, in vivo pull-down from mouse retinas, and proximity ligation assay to identify and confirm the SNARE proteins Syntaxin 3B and SNAP-25 as novel binding partners of RDS and ROM-1. We show that both covalently linked and non-covalently linked RDS complexes interact with Syntaxin 3B. RDS in the mouse is trafficked from the inner segment to the outer segment by both conventional (i.e., Golgi dependent) and unconventional secretory pathways, and RDS from both pathways interacts with Syntaxin3B. Syntaxin 3B and SNAP-25 are enriched in the inner segment (compared to the outer segment) suggesting that the interaction with RDS/ROM-1 occurs in the inner segment. Syntaxin 3B and SNAP-25 are involved in mediating fusion of vesicles carrying other outer segment proteins during outer segment targeting, so could be involved in the trafficking of RDS/ROM-1. PMID:26406599

  1. 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.

  2. Fuel pins with both target and fuel pellets in an isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

  3. Assemblies with both target and fuel pins in an isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

  4. Vented target elements for use in an isotope-production reactor. [LMFBR

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium gas in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins equipped with vents, and tritium gas is recovered from the coolant.

  5. A Nodal Kinetics and Thermohydraulics Analysis (NOKTA) Code for Analyzing Rod-Ejection Accidents and Other Transients in Nuclear Power Reactor Cores

    SciTech Connect

    Kaya, Sadi; Yavuz, Hasbi

    2000-01-15

    For analyzing nuclear power reactor core transients, a three-dimensional nodal kinetics and thermohydraulics code, NOKTA, was developed. Nodal kinetics calculation is based on a one-group neutron diffusion approach. Thermal-hydraulics analysis is handled as in the COBRA-IV-I code. The NOKTA code was designed for analyzing especially large reactivity accidents, such as sudden rod ejection. It can also analyze intermediate transients, such as sharp power changes that may initiate xenon oscillations, and slow transients, such as boric acid density changes in the flow. The code dimensions are set at 125 subchannels and 30 axial levels. Calculation starts with a saturated xenon density, one-group neutronics parameters, and a flux profile, which is required as an input. Initially, k{sub eff} of each computation cell is set to unity.

  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. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    SciTech Connect

    Shott, Gregory

    2014-08-31

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  8. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    SciTech Connect

    Samuel Bays; Pavel Medvedev; Michael Pope; Rodolfo Ferrer; Benoit Forget; Mehdi Asgari

    2009-04-01

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  9. NEUTRONIC REACTORS

    DOEpatents

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  10. 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.

  11. Parametric study of spallation targets for the MYRRHA reactor using MCNPX simulations

    NASA Astrophysics Data System (ADS)

    Rebello, A. L. P.; Martinez, A. S.; Gonçalves, A. C.

    2014-06-01

    The present work aims to evaluate the behavior of neutron multiplicity in a spallation target using MCNPX simulations, focusing on its application in the MYRRHA reactor. It was studied the two types of spallation target proposed for the MYRRHA project, windowless and windows target, in order to compare them and find saturation boundaries. Some saturation boundaries were found and the windowless target proved to be as viable as the windows one. Each one produced nearly the same number of neutrons per incident proton. Using the concept of neutron cost, it was also observed that the optimum conditions on neutron production occur at about 1GeV, for both target designs.

  12. 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)

  13. 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.

  14. 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)

  15. Systems and methods for processing irradiation targets through a nuclear reactor

    DOEpatents

    Dayal, Yogeshwar; Saito, Earl F.; Berger, John F.; Brittingham, Martin W.; Morales, Stephen K.; Hare, Jeffrey M.

    2016-05-03

    Apparatuses and methods produce radioisotopes in instrumentation tubes of operating commercial nuclear reactors. Irradiation targets may be inserted and removed from instrumentation tubes during operation and converted to radioisotopes otherwise unavailable during operation of commercial nuclear reactors. Example apparatuses may continuously insert, remove, and store irradiation targets to be converted to useable radioisotopes or other desired materials at several different origin and termination points accessible outside an access barrier such as a containment building, drywell wall, or other access restriction preventing access to instrumentation tubes during operation of the nuclear plant.

  16. NEUTRONIC REACTOR MANIPULATING DEVICE

    DOEpatents

    Ohlinger, L.A.

    1962-08-01

    A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

  17. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  18. Neutronic Assessment of Transmutation Target Compositions in Heterogeneous Sodium Fast Reactor Geometries

    SciTech Connect

    Samuel E. Bays; Rodolfo M. Ferrer; Michael A. Pope; Benoit Forget; Mehdi Asgari

    2008-02-01

    The sodium fast reactor is under consideration for consuming the transuranic waste in the spent nuclear fuel generated by light water reactors. This work is concerned with specialized target assemblies for an oxide-fueled sodium fast reactor that are designed exclusively for burning the americium and higher mass actinide component of light water reactor spent nuclear fuel (SNF). The associated gamma and neutron radioactivity, as well as thermal heat, associated with decay of these actinides may significantly complicate fuel handling and fabrication of recycled fast reactor fuel. The objective of using targets is to isolate in a smaller number of assemblies these concentrations of higher actinides, thus reducing the volume of fuel having more rigorous handling requirements or a more complicated fabrication process. This is in contrast to homogeneous recycle where all recycled actinides are distributed among all fuel assemblies. Several heterogeneous core geometries were evaluated to determine the fewest target assemblies required to burn these actinides without violating a set of established fuel performance criteria. The DIF3D/REBUS code from Argonne National Laboratory was used to perform the core physics and accompanying fuel cycle calculations in support of this work. Using the REBUS code, each core design was evaluated at the equilibrium cycle condition.

  19. Study on Fracture Behavior of 2D-C/C Composite for Application to Control Rod of Very High Temperature Reactor

    NASA Astrophysics Data System (ADS)

    Sumita, J.; Fujita, I.; Shibata, T.; Makita, T.; Takagi, T.; Kunimoto, E.; Sawa, K.; Kim, W.; Park, J.

    2011-10-01

    For a control rod element of the Very High Temperature Reactor, a carbon fiber reinforced carbon matrix composite (C/C composite) is one of the major candidate materials for its high strength and thermal stability. In this study, in order to establish the data base of the 2D-C/C composite, the fracture data was obtained by simulating the crack expected to be generated under the VHTR condition and the oxidation effect on the fracture behavior was evaluated. Moreover, the fracture mechanism of the C/C composite was investigated through scanning electron microscope observation. This study showed that the oxidized matrix caused reduction of the fracture toughness and the reduction ratio was dependent on the density of matrix and a number cracks. With increasing the oxidation, the fracture toughness is mainly dependent on the fiber characteristics. Furthermore, the crack grows along the boundary between fiber bundles without breaking the fiber. The cracks which were initiated at the interface between the matrix and the fiber were gathered into the voids in the boundary between fiber bundles, and, then, the cracks grew up in the matrix.

  20. Production of {sup 99}Mo using LEU and molybdenum targets in a 1 MW Triga reactor

    SciTech Connect

    Mo, S.C.

    1993-12-31

    The production of {sup 99}Mo using Low Enriched Uranium (LEU) and natural molybdenum targets in a 1 MW Triga reactor is investigated. The successive linear programming technique is applied to minimize the target loadings for different yield constraints. The irradiation time is related to the kinetics of the growth and decay of {sup 99}Mo. The feasibility of a neutron generated based {sup 99}Mo production system is discussed.

  1. 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.

  2. 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.

  3. Feasibility study Part I - Thermal hydraulic analysis of LEU target for {sup 99}Mo production in Tajoura reactor

    SciTech Connect

    Bsebsu, F.M.; Abotweirat, F. E-mail: abutweirat@yahoo.com; Elwaer, S.

    2008-07-15

    The Renewable Energies and Water Desalination Research Center (REWDRC), Libya, will implement the technology for {sup 99}Mo isotope production using LEU foil target, to obtain new revenue streams for the Tajoura nuclear research reactor and desiring to serve the Libyan hospitals by providing the medical radioisotopes. Design information is presented for LEU target with irradiation device and irradiation Beryllium (Be) unit in the Tajoura reactor core. Calculated results for the reactor core with LEU target at different level of power are presented for steady state and several reactivity induced accident situations. This paper will present the steady state thermal hydraulic design and transient analysis of Tajoura reactor was loaded with LEU foil target for {sup 99}Mo production. The results of these calculations show that the reactor with LEU target during the several cases of transient are in safe and no problems will occur. (author)

  4. 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.

  5. 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.

  6. Method for depleting BWRs using optimal control rod patterns

    SciTech Connect

    Taner, M.S.; Levine, S.H. ); Hsiao, M.Y. )

    1991-01-01

    Control rod (CR) programming is an essential core management activity for boiling water reactors (BWRs). After establishing a core reload design for a BWR, CR programming is performed to develop a sequence of exposure-dependent CR patterns that assure the safe and effective depletion of the core through a reactor cycle. A time-variant target power distribution approach has been assumed in this study. The authors have developed OCTOPUS to implement a new two-step method for designing semioptimal CR programs for BWRs. The optimization procedure of OCTOPUS is based on the method of approximation programming and uses the SIMULATE-E code for nucleonics calculations.

  7. 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.

  8. Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

    DOEpatents

    Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.

    2015-12-29

    Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

  9. 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.

  10. NEUTRONIC REACTOR

    DOEpatents

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  11. CONTROL FOR NEUTRONIC REACTOR

    DOEpatents

    Lichtenberger, H.V.; Cameron, R.A.

    1959-03-31

    S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

  12. 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.

  13. A target station for plasma exposure of neutron irradiated fusion material samples to reactor relevant conditions

    NASA Astrophysics Data System (ADS)

    Rapp, Juergen; Giuliano, Dominic; Ellis, Ronald; Howard, Richard; Lore, Jeremy; Lumsdaine, Arnold; Lessard, Timothy; McGinnis, William; Meitner, Steven; Owen, Larry; Varma, Venugopal

    2015-11-01

    The Material Plasma Exposure eXperiment (MPEX) is a device planned to address scientific and technological gaps for the development of viable plasma facing components for fusion reactor conditions (FNSF, DEMO). It will have to address the relevant plasma conditions in a reactor divertor (electron density, electron temperature, ion fluxes) and it needs to be able to expose a-priori neutron irradiated samples. A pre design of a target station able to handle activated materials will be presented. This includes detailed MCNP as well as SCALE and MAVRIC calculations for all potential plasma-facing materials to estimate dose rates. Details on the remote handling schemes for the material samples will be presented. 2 point modeling of the linear plasma transport has been used to scope out the parameter range of the anticipated power fluxes to the target. This has been used to design the cooling capability of the target. The operational conditions of surface temperatures, plasma conditions, and oblique angle of incidence of magnetic field to target surface will be discussed. ORNL is managed by UT-Battelle, LLC, for the U.S. DOE under contract DE-AC-05-00OR22725.

  14. Nuclear reactor control apparatus

    DOEpatents

    Sridhar, Bettadapur N.

    1983-11-01

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  15. NUCLEAR REACTOR

    DOEpatents

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

  16. Fast-acting nuclear reactor control device

    DOEpatents

    Kotlyar, Oleg M.; West, Phillip B.

    1993-01-01

    A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

  17. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    DOE PAGESBeta

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; et al

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  18. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    SciTech Connect

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; Fulwyler, James

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  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. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  1. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  2. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  3. Fast reactor core concepts to improve transmutation efficiency

    SciTech Connect

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-12-31

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

  4. 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

  5. Linear motion device and method for inserting and withdrawing control rods

    DOEpatents

    Smith, Jay E.

    1984-01-01

    A linear motion device, more specifically a control rod drive mechanism (CRDM) for inserting and withdrawing control rods into a reactor core, is capable of independently and sequentially positioning two sets of control rods with a single motor stator and rotor. The CRDM disclosed can control more than one control rod lead screw without incurring a substantial increase in the size of the mechanism.

  6. Systems and methods for retaining and removing irradiation targets in a nuclear reactor

    DOEpatents

    Runkle, Gary A.; Matsumoto, Jack T.; Dayal, Yogeshwar; Heinold, Mark R.

    2015-12-08

    A retainer is placed on a conduit to control movement of objects within the conduit in access-restricted areas. Retainers can prevent or allow movement in the conduit in a discriminatory fashion. A fork with variable-spacing between prongs can be a retainer and be extended or collapsed with respect to the conduit to change the size of the conduit. Different objects of different sizes may thus react to the fork differently, some passing and some being blocked. Retainers can be installed in inaccessible areas and allow selective movement in remote portions of conduit where users cannot directly interface, including below nuclear reactors. Position detectors can monitor the movement of objects through the conduit remotely as well, permitting engagement of a desired level of restriction and object movement. Retainers are useable in a variety of nuclear power plants and with irradiation target delivery, harvesting, driving, and other remote handling or robotic systems.

  7. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  8. 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.

  9. REACTOR CONTROL

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  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. Fission control system for nuclear reactor

    DOEpatents

    Conley, G.H.; Estes, G.P.

    Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

  12. The Application of Long Esr Sensor Rods for Neutron and Gamma Dosimetry of the "weak" In-Reactor Irradiation of the Htgr Fuel

    NASA Astrophysics Data System (ADS)

    Usatyi, A. F.; Momot, G. V.; Kaynov, V. B.; Kuznetsov, A. I.

    2003-06-01

    In order to measure the general spatial distribution of the thermal neutron fluence during the so called "weak" irradiation (less than 1017 n/m2) of HTGR nuclear fuel for subsequent high temperature tests including fission products release, we apply local (0.3 cm rings) and distributed (long rods up to 65 cm) accumulative detectors of neutrons and gamma with results' reading by the electron spin resonance method (ESR-sensors). Sensors materials are: silicate ceramic (glass) containing B2O3 (neutron sensor) and quartz with Al2O3 addition (gamma sensor). The new possibilities of nontraditional ESR-sensors, a new type of nuclear radiation detectors are discussed.

  13. 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.

  14. 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.

  15. Safety evaluation report related to the Department of Energy`s proposal for the irradiation of lead test assemblies containing tritium-producing burnable absorber rods in commercial light-water reactors. Project Number 697

    SciTech Connect

    1997-05-01

    The NRC staff has reviewed a report, submitted by DOE to determine whether the use of a commercial light-water reactor (CLWR) to irradiate a limited number of tritium-producing burnable absorber rods (TPBARs) in lead test assemblies (LTAs) raises generic issues involving an unreviewed safety question. The staff has prepared this safety evaluation to address the acceptability of these LTAs in accordance with the provision of 10 CFR 50.59 without NRC licensing action. As summarized in Section 10 of this safety evaluation, the staff has identified issues that require NRC review. The staff has also identified a number of areas in which an individual licensee undertaking irradiation of TPBAR LTAs will have to supplement the information in the DOE report before the staff can determine whether the proposed irradiation is acceptable at a particular facility. The staff concludes that a licensee undertaking irradiation of TPBAR LTAs in a CLWR will have to submit an application for amendment to its facility operating license before inserting the LTAs into the reactor.

  16. Advances in implosion physics, alternative targets design, and neutron effects on heavy ion fusion reactors

    NASA Astrophysics Data System (ADS)

    Velarde, G.; Perlado, J. M.; Alonso, E.; Alonso, M.; Domínguez, E.; Rubiano, J. G.; Gil, J. M.; Gómez del Rio, J.; Lodi, D.; Malerba, L.; Marian, J.; Martel, P.; Martínez-Val, J. M.; Mínguez, E.; Piera, M.; Ogando, F.; Reyes, S.; Salvador, M.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P.

    2001-05-01

    The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DT x fuel with a small tritium initial content ( x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures (≫100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower Te and to enhance radiation losses, reducing the plasma temperature, Ti. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination barriers in Si

  17. 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.

  18. NUCLEAR REACTOR

    DOEpatents

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  19. SPRING DRIVEN ACTUATING MECHANISM FOR NUCLEAR REACTOR CONTROL

    DOEpatents

    Bevilacqua, F.; Uecker, D.F.; Groh, E.F.

    1962-01-23

    l962. rod in a nuclear reactor to shut it down. The control rod or an extension thereof is wound on a drum as it is withdrawn from the reactor. When an emergency occurs requiring the reactor to be shut down, the drum is released so as to be free to rotate, and the tendency of the control rod or its extension coiled on the drum to straighten itself is used for quickly returning the control rod to the reactor. (AEC)

  20. Neutron flux characterization of a peripheral target position in the High Flux Isotope Reactor.

    PubMed

    Garland, M A; Mirzadeh, S; Alexander, C W; Hirtz, G J; Hobbs, R W; Pertmer, G A; Knapp, F F

    2003-07-01

    The High Flux Isotope Reactor at the Oak Ridge National Laboratory provides the highest steady-state thermal neutron flux in the western world for a wide range of experiments and for isotope production. The highest available fluxes are located in a flux trap region created inside the nested fuel elements. The experimentally determined thermal and the empirically obtained epithermal flux values along the vertical axis of the peripheral target position were fit to cosine curves, with the thermal flux ranging from 1.1 x 10(15)ns(-1)cm(-2) at outer positions to 1.5 x 10(15)ns(-1)cm(-2) at the center. The corresponding epithermal flux ranged from 3.5 x 10(13) to 7.5 x 10(13)ns(-1)cm(-2), respectively. The fast neutron flux (En > or = 0.32 MeV in two positions and En > or = 1.5 MeV in two other positions) was approximately 6 x 10(14)ns(-1)cm(-2), corresponding to a fast to thermal ratio of approximately 0.4.

  1. 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.

  2. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    SciTech Connect

    PARMA JR.,EDWARD J.

    2000-01-01

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

  3. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  4. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1962-12-25

    A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

  5. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  6. High solids fermentation reactor

    DOEpatents

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  7. High solids fermentation reactor

    DOEpatents

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  8. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  9. CONTROL SYSTEM FOR NEUTRONIC REACTORS

    DOEpatents

    Crever, F.E.

    1962-05-01

    BS>A slow-acting shim rod for control of major variations in reactor neutron flux and a fast-acting control rod to correct minor flux variations are employed to provide a sensitive, accurate control system. The fast-acting rod is responsive to an error signal which is produced by changes in the neutron flux from a predetermined optimum level. When the fast rod is thus actuated in a given direction, means is provided to actuate the slow-moving rod in that direction to return the fast rod to a position near the midpoint of its control range. (AEC)

  10. 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.

  11. 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.

  12. 1987 Sucker rod tables

    SciTech Connect

    Not Available

    1987-03-01

    This reference identifies manufacturers qualified to produce API sucker rods and related equipment, lists chemical and mechanical properties of the various types of rods and provides dimensional characteristics. In addition, similar information is given for non-API products such as fiberglass and hollow rods.

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

    NASA Astrophysics Data System (ADS)

    Seiler, J. M.; Rameau, B.

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

  14. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

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

    1963-06-11

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

  15. NEUTRONIC REACTOR STRUCTURE

    DOEpatents

    Daniels, F.

    1961-10-24

    A reactor core, comprised of vertical stacks of hexagonal blocks of beryllium oxide having axial cylindrical apertures extending therethrough and cylindrical rods of a sintered mixture of uranium dioxide and beryllium oxide, is described. (AEC)

  16. Evaluation of Heterogeneous Options: Effects of MgO versus UO2 Matrix Selection for Minor Actinide Targets in a Sodium Fast Reactor

    SciTech Connect

    M. Pope; S. Bays; R. Ferrer

    2008-03-01

    The primary focus of this work was to compare MgO with UO2 as target matrix material options for burning minor actinides in a transmutation target within a sodium fast reactor. This analysis compared the transmutation performance of target assemblies having UO2 matrix to those having specifically MgO inert matrix.

  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. Targeted Ablation of the Pde6h Gene in Mice Reveals Cross-species Differences in Cone and Rod Phototransduction Protein Isoform Inventory*

    PubMed Central

    Brennenstuhl, Christina; Tanimoto, Naoyuki; Burkard, Markus; Wagner, Rebecca; Bolz, Sylvia; Trifunovic, Dragana; Kabagema-Bilan, Clement; Paquet-Durand, Francois; Beck, Susanne C.; Huber, Gesine; Seeliger, Mathias W.; Ruth, Peter; Wissinger, Bernd; Lukowski, Robert

    2015-01-01

    Phosphodiesterase-6 (PDE6) is a multisubunit enzyme that plays a key role in the visual transduction cascade in rod and cone photoreceptors. Each type of photoreceptor utilizes discrete catalytic and inhibitory PDE6 subunits to fulfill its physiological tasks, i.e. the degradation of cyclic guanosine-3′,5′-monophosphate at specifically tuned rates and kinetics. Recently, the human PDE6H gene was identified as a novel locus for autosomal recessive (incomplete) color blindness. However, the three different classes of cones were not affected to the same extent. Short wave cone function was more preserved than middle and long wave cone function indicating that some basic regulation of the PDE6 multisubunit enzyme was maintained albeit by a unknown mechanism. To study normal and disease-related functions of cone Pde6h in vivo, we generated Pde6h knock-out (Pde6h−/−) mice. Expression of PDE6H in murine eyes was restricted to both outer segments and synaptic terminals of short and long/middle cone photoreceptors, whereas Pde6h−/− retinae remained PDE6H-negative. Combined in vivo assessment of retinal morphology with histomorphological analyses revealed a normal overall integrity of the retinal organization and an unaltered distribution of the different cone photoreceptor subtypes upon Pde6h ablation. In contrast to human patients, our electroretinographic examinations of Pde6h−/− mice suggest no defects in cone/rod-driven retinal signaling and therefore preserved visual functions. To this end, we were able to demonstrate the presence of rod PDE6G in cones indicating functional substitution of PDE6. The disparities between human and murine phenotypes caused by mutant Pde6h/PDE6H suggest species-to-species differences in the vulnerability of biochemical and neurosensory pathways of the visual signal transduction system. PMID:25739440

  19. Targeted ablation of the Pde6h gene in mice reveals cross-species differences in cone and rod phototransduction protein isoform inventory.

    PubMed

    Brennenstuhl, Christina; Tanimoto, Naoyuki; Burkard, Markus; Wagner, Rebecca; Bolz, Sylvia; Trifunovic, Dragana; Kabagema-Bilan, Clement; Paquet-Durand, Francois; Beck, Susanne C; Huber, Gesine; Seeliger, Mathias W; Ruth, Peter; Wissinger, Bernd; Lukowski, Robert

    2015-04-17

    Phosphodiesterase-6 (PDE6) is a multisubunit enzyme that plays a key role in the visual transduction cascade in rod and cone photoreceptors. Each type of photoreceptor utilizes discrete catalytic and inhibitory PDE6 subunits to fulfill its physiological tasks, i.e. the degradation of cyclic guanosine-3',5'-monophosphate at specifically tuned rates and kinetics. Recently, the human PDE6H gene was identified as a novel locus for autosomal recessive (incomplete) color blindness. However, the three different classes of cones were not affected to the same extent. Short wave cone function was more preserved than middle and long wave cone function indicating that some basic regulation of the PDE6 multisubunit enzyme was maintained albeit by a unknown mechanism. To study normal and disease-related functions of cone Pde6h in vivo, we generated Pde6h knock-out (Pde6h(-/-)) mice. Expression of PDE6H in murine eyes was restricted to both outer segments and synaptic terminals of short and long/middle cone photoreceptors, whereas Pde6h(-/-) retinae remained PDE6H-negative. Combined in vivo assessment of retinal morphology with histomorphological analyses revealed a normal overall integrity of the retinal organization and an unaltered distribution of the different cone photoreceptor subtypes upon Pde6h ablation. In contrast to human patients, our electroretinographic examinations of Pde6h(-/-) mice suggest no defects in cone/rod-driven retinal signaling and therefore preserved visual functions. To this end, we were able to demonstrate the presence of rod PDE6G in cones indicating functional substitution of PDE6. The disparities between human and murine phenotypes caused by mutant Pde6h/PDE6H suggest species-to-species differences in the vulnerability of biochemical and neurosensory pathways of the visual signal transduction system. PMID:25739440

  20. Sucker rod construction

    SciTech Connect

    Anderson, R.A.; Goodman, J.L.; Tickle, J.D.; Liskey, A.K.

    1987-03-31

    A sucker rod construction is described comprising: a connector member being formed to define a rod receptacle having a closed axially inner end and an open axially outer end, the rod receptacle having axially spaced, tapered annular surfaces, a cylindrical fiberglass rod having an end having an outer surface being received within the rod receptacle through the outer end and cooperating therewith to define an annular chamber between the outer surface of the end of the rod and the tapered annular surfaces, and a bonding means positioned in the annular chamber for bonding to the outer surface of the end of the rod to confront the tapered annular surfaces, each annular surface having an angle of taper with respect to the outer surface of the fiberglass rod, and each angle of taper being progressively and uniformly less toward the open end by an amount between one and one-half degrees and two degrees, inclusive, and a collet connected to the connector member adjacent the open axially outer end of the rod receptacle and having an axial bore therethrough retaining the end of the rod in coaxial position within the rod receptacle.

  1. Wear resistant rod guide

    SciTech Connect

    Gray, K.W.

    1991-12-03

    This paper describes a sucker rod guide. It comprises: a series of sucker rods connected end to end forming a sucker rod string, the sucker rod string extending down into a tubing string of a producing oil well from a pump jack located on the surface of the ground above the tubing string to a pump located at a bottom end of the tubing string, the pump forces produced fluid collected at the bottom end of the tubing string up to the ground's surface, the produced fluid occupies a space between the rod string and the tubing string through which the fluid is channeled from the bottom end of the tubing string to the ground's surface, the pump jack raises and lowers the rod string in the fluid being pumped up the tubing string while the fluid bathes the rod string within the tubing string, wherein the improvement comprises the following structure in combination with the above.

  2. Sucker rod guide

    SciTech Connect

    Edwards, B.J.; Starks, J.A.

    1989-08-22

    This patent describes a sucker rod guide for mounting on a sucker rod and spacing the sucker rod from the tubing in an oil well. The guide comprising a generally cylindrically-shaped, extruded, ultra-high density polyethylene body having a substantially smooth outside surface; a longitudinal bore provided centrally of the body. The bore having a smaller diameter than the diameter of the sucker rod; a plurality of grooves provided in circumferential relationship in the bore; and a tapered slot extending longitudinally through the body from the outside surface to the bore. The tapered slot further comprising a slot mouth located at the outside surface and a slot throat spaced from the slot mouth. The slot throat lying adjacent to the sucker rod bore and wherein the slot throat is wider than the slot mouth for mounting the sucker rod guide on the sucker rod.

  3. Low turbulence rod guide

    SciTech Connect

    Olinger, E.L.

    1992-05-26

    This patent describes an improved sucker rod guide for fixedly engaging around a sucker rod at a selected location along the length of the rod. It comprises a substantially cylindrical polymeric body having a longitudinal axis, a terminal end substantially continually tapered to the rod, a radially-inward surface and a radially outward surface, the radially inward surface of the body adjacent to and in tripping engagement with the rod when the rod guide is fixedly engaged around the rod; and a plurality of substantially continuous, longitudinal vanes carried by the body, a vane having a selected length and width, and longitudinally disposed along the radially outward surface of the guide body, extending radially away from the guide body and having a radially outside wear surface.

  4. 26. A typical outer rod room, or rack room, showing ...

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

    26. A typical outer rod room, or rack room, showing the racks for the nine horizontal control rods (HCRs) that would be inserted or withdrawn from the pile to control the rate of reaction. In this case, it is the 105-F Reactor in February 1945. The view is looking away from the pile, which is out of the picture on the left. Several of the cooling water hose reels for the rods can be seen at the end of the racks near the wall. D-8323 - B Reactor, Richland, Benton County, WA

  5. The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

    NASA Astrophysics Data System (ADS)

    Scarlat, Raluca O.; Peterson, Per F.

    2014-01-01

    The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

  6. Hydraulic lock for displacer rod drive mechanism (DRDM) and method of operation

    SciTech Connect

    Rinker, E.D.

    1990-12-18

    This paper describes a drive rod latch in combination with a nuclear reactor having a drive rod disposed in a rod housing characterized in that the drive rod has one end selectively exposed to a first, relatively low pressure zone of the reactor and another end thereof in communication with a second, relatively high pressure zone of the reactor. The drive rod further having disposed on an end thereof a valve member and the rod housing having disposed thereon a corresponding valve seat, and a control valve for selectively establishing communication between the housing and the first zone of the reactor whereby a pressure differential is created across the piston. The pressure differential being sufficient to seat the valve member against the valve seat to thereby establish a pressure boundary.

  7. POWER BREEDER REACTOR

    DOEpatents

    Monson, H.O.

    1960-11-22

    An arrangement is offered for preventing or minimizing the contraction due to temperature rise, of a reactor core comprising vertical fuel rods in sodium. Temperature rise of the fuel rods would normally make them move closer together by inward bowing, with a resultant undesired increase in reactivity. According to the present invention, assemblies of the fuel rods are laterally restrained at the lower ends of their lower blanket sections and just above the middle of the fuel sections proper of the rods, and thus the fuel sections move apart, rather than together, with increase in temperature.

  8. Vibration of the Package of Rods Linked by Spacer Grids

    NASA Astrophysics Data System (ADS)

    Zeman, V.; Hlaváč, Z.

    This paper deals with modelling and vibration analysis of the large package of identical parallel rods which are linked by transverse springs (spacer grids) placed on several level spacings. The vibration of rods is caused by the support plate motion. The rod discretization by FEM is based on Rayleigh beam theory. With respect to cyclic and central package rod symmetry, the system is decomposed to identical revolved rod segments. The modal synthesis method with condensation of the rod segments is used for modelling and determination of steady forced vibration of the whole system. The presented method is the first step to modelling of the nuclear fuel assembly vibration caused by kinematical excitation determined by motion of the support plates which are part of the reactor core.

  9. Type B investigation of the iridium contamination event at the High Flux Isotope Reactor on September 7, 1993

    SciTech Connect

    Not Available

    1994-03-01

    On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci{sup 192}Ir) from the HFIR pool side to a transport truck. Small amounts of Ir were released from the cask onto the reactor bay floor. The floor was cleaned, and the cask was shipped to a hot cell at Building 3047 on Oct. 3, 1993. The event was caused by rupture of one of the Ir target rods after it was loaded into the cask for normal transport operations; the rupture was the result of steam generation in the target rod soon after it was placed in the cask (water had entered the target rod through a tiny defect in a weld while it was in the reactor under pressure). While the target rods were in the reactor and reactor pool, there was sufficient cooling to prevent steam generation; when the target rod was loaded into the dry transport cask, the temperature increased enough to result in boiling of the trapped water and produced high enough pressure to result in rupture. The escaping steam ejected some of the Ir pellets. The event was reported as Occurrence Report Number ORO--MMES-X10HFIR-1993-0030, dated Sept. 8, 1993. Analysis indicated that the following conditions were probable causes: less than adequate welding procedures, practices, or techniques, material controls, or inspection methods, or combination thereof, could have led to weld defects, affecting the integrity of target rod IR-75; less than adequate secondary containment in the cask allowed Ir pellets to escape.

  10. 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.

  11. Pull rod assembly

    DOEpatents

    Cioletti, Olisse C.

    1990-01-01

    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.

  12. Pull rod assembly

    SciTech Connect

    Cioletti, O.C.

    1990-05-22

    This patent describes a pull rod assembly. It comprises: a pull rod having three peripheral grooves, a piston device including an adaptor ring and a seal ring. The piston device is 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.

  13. Linear motion device and method for inserting and withdrawing control rods

    DOEpatents

    Smith, J.E.

    Disclosed is a linear motion device and more specifically a control rod drive mechanism (CRDM) for inserting and withdrawing control rods into a reactor core. The CRDM and method disclosed is capable of independently and sequentially positioning two sets of control rods with a single motor stator and rotor. The CRDM disclosed can control more than one control rod lead screw without incurring a substantial increase in the size of the mechanism.

  14. HTCAP-1: a program for calcuating operating temperatures in HFIR target irradiation experiments

    SciTech Connect

    Kania, M.J.; Howard A.M.

    1980-06-01

    The thermal modeling code, HTCAP-1, calculates in-reactor operating temperatures of fueled specimens contained in the High Flux Isotope Reactor (HFIR) target irradiation experiments (HT-series). Temperature calculations are made for loose particle and bonded fuel rod specimens. Maximum particle surface temperatures are calculated for the loose particles and centerline and surface temperatures for the fuel rods. Three computational models are employed to determine fission heat generation rates, capsule heat transfer analysis, and specimen temperatures. This report is also intended to be a users' manual, and the application of HTCAP-1 to the HT-34 irradiation capsule is presented.

  15. Analysis of a Partial MOX Core Design with Tritium Targets for Light Water Reactors

    SciTech Connect

    Anistratov, Dmitriy Y.; Adams, Marvin L.

    1998-04-19

    This report constitutes tangible and verifiable deliverable associated with the task To study the effects of using WG MOX fuel in tritium-producing LWR” of the subproject Water Reactor Options for Disposition of Plutonium. The principal investigators of this subproject are Naeem M. Abdurrahman of the University of Texas at Austin and Marvin L. Adams of Texas A&M University. This work was sponsored by the Amarillo National Resource Center for Plutonium.

  16. Nuclear reactor control apparatus

    DOEpatents

    Sridhar, Bettadapur N.

    1983-10-25

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

  17. Reactivity control assembly for nuclear reactor

    DOEpatents

    Bollinger, Lawrence R.

    1984-01-01

    Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

  18. 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.

  19. Rat hippocampal neurons express genes for both rod retinal and olfactory cyclic nucleotide-gated channels: novel targets for cAMP/cGMP function.

    PubMed Central

    Kingston, P A; Zufall, F; Barnstable, C J

    1996-01-01

    Cyclic nucleotide-gated (CNG) channels are Ca(2+)-permeable, nonspecific cation channels that can be activated through direct interaction with cAMP and/or cGMP. Recent electrophysiological evidence for these channels in cultured hippocampal neurons prompted us to investigate the expression of CNG channel genes in hippocampus. PCR amplification detected the expression of transcripts for subunit 1 of both the rod photoreceptor (RCNGC1) and the olfactory receptor cell (OCNGC1) subtype of CNG channel in adult rat hippocampus. In situ hybridization detected expression of both channel subtypes in most principal neurons, including pyramidal cells of the CA1 through CA3 regions and granule cells of the dentate gyrus. From the hybridization patterns, we conclude that the two genes are colocalized in individual neurons. Comparison of the patterns of expression of type 1 cGMP-dependent protein kinase and the CNG channels suggests that hippocampal neurons can respond to changes in cGMP levels with both rapid changes in CNG channel activity and slower changes induced by phosphorylation. Future models of hippocampal function should include CNG channels and their effects on both electrical responses and intracellular Ca2+ levels. Images Fig. 1 Fig. 2 Fig. 3 Fig. 4 Fig. 5 Fig. 6 PMID:8816819

  20. 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)

  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. 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.

  3. Safety control circuit for a neutronic reactor

    DOEpatents

    Ellsworth, Howard C.

    2004-04-27

    A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

  4. 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.

  5. Molybdenum-99 production from reactor irradiation of molybdenum targets: a viable strategy for enhanced availability of technetium-99m.

    PubMed

    Pillai, M R A; Knapp, F F Russ

    2012-08-01

    Fission-produced 99Mo (F 99Mo) is traditionally used for fabrication of 99Mo/99mTc alumina-based column generators. In this paper, several emerging strategies are discussed which are being pursued or have been suggested to overcome the continuing shortages of F 99Mo. In addition to the hopeful eventual success of these proposed new 99Mo and 99mTc production technologies, an additional attractive strategy is the alternative production and use of low specific activity (LSA) 99Mo. This strategy avoids fission and is accomplished by direct activation of molybdenum targets in nuclear reactors, which would preclude sole continued reliance on F 99Mo. The principal focus of this paper is a detailed discussion on the advantages and strategies for enhanced production of LSA 99Mo using an international network of research reactors. Several effective strategies are discussed to obtain 99mTc from LSA 99Mo as well as more efficient use of the alumina-based generator system. The delayed time period between 99Mo production and traditional 99Mo/99mTc alumina column generator manufacture and distribution to user sites results in the loss of more than 50% of 99Mo activity. Another strategy is a paradigm shift in the use of 99Mo by recovering clinical-grade 99mTc from 99Mo solution as an alternative to use of 99Mo/99mTc column generators, thereby avoiding substantial decreased availability of 99Mo from radioactive decay. Implementation of the suggested strategies would be expected to increase availability of 99mTc to the clinical user community by several fold. Additional important advantages for the use of LSA 99Mo include eliminating the need for fission product waste management and precluding proliferation concerns by phasing out the need for high (HEU)- and low (LEU)-enriched uranium targets required for F 99Mo production.

  6. Why rods and cones?

    PubMed

    Lamb, T D

    2016-02-01

    Under twenty-first-century metropolitan conditions, almost all of our vision is mediated by cones and the photopic system, yet cones make up barely 5% of our retinal photoreceptors. This paper looks at reasons why we additionally possess rods and a scotopic system, and asks why rods comprise 95% of our retinal photoreceptors. It considers the ability of rods to reliably signal the arrival of individual photons of light, as well as the ability of the retina to process these single-photon signals, and it discusses the advantages that accrue. Drawbacks in the arrangement, including the very slow dark adaptation of scotopic vision, are also considered. Finally, the timing of the evolution of cone and rod photoreceptors, the retina, and the camera-style eye is summarised.

  7. Long-Rod Moving-Plate Interaction

    NASA Astrophysics Data System (ADS)

    Partom, Y.

    2002-07-01

    Understanding the mechanics of interaction of a long rod projectile with a forward moving plate at an angle is essential to understanding long rod interaction with an explosive reactive armor cassette. To investigate the mechanics of such an interaction we use AUTODIN2D/EULER in plane geometry, although the problem is 3D. We assume that this is a satisfactory approximation, as we're only interested in the main features, and are not comparing fine details to experimental results. From the simulations we learn that the interaction never reaches steady state. Initially each material splits into two streams, and the interaction plane is perpendicular to the rod. But with time the interaction plane rotates slowly, until it becomes parallel to the rod, which is then able to continue moving forward without interruption. During this process interacting rod material of length DeltaL is diverted at an angle and becomes ineffective for penetrating the main target. We made many such runs to determine the dependence of DeltaL on the parameters of the problem. This dependence makes it possible to predict DeltaL for a variety of rod-plate situations.

  8. Nuclear reactor safety device

    DOEpatents

    Hutter, E.

    1983-08-15

    A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

  9. Nuclear reactor shutdown system

    DOEpatents

    Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  10. Flexible sucker rod unit

    SciTech Connect

    Allen, L.F.

    1987-02-03

    This patent describes a deep well having: a. an education tube with an inside diameter extending from the surface of the earth to far below the surface, b. a reciprocating pump housing attached to the bottom of the education tube, c. pump jack means at the surface for reciprocating the pump, d. a light sucker rod connected to the pump jack means and extending into the education tube, and e. a series of heavy sinker bars having a large cross sectional area in the education tube connecting the light sucker rod to the pump; f. an improved integral metal flexible rod unit interconnecting the sinker bars comprising in combination with the above: g. a coupling on each end of the integral metal flexible rod unit connecting the flexible rod unit to the contiguous sinker bar, h. a segment which is flexible as compared to the sinker bars connecting one of the couplings to i. an integral metal bearing adjacent to the other of the couplings, the bearing having j. a cylindrical surface with k. a diameter i. only slightly smaller than the inside diameter of the education tube thereby forming a sliding fit therewith, and ii. greater than the diameter of any other portion of the flexible rod unit and the sinker bar, and l. grooves in the cylindrical surface for the passage of fluid between in the education tube around the bearing.

  11. Advanced light water reactor requirements document: Chapter 4, Reactor systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the following for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR): reactor pressure vessel, nozzles and safe-ends, reactor internals, in-vessel portions of fluid systems (including reactor internal pumps (Emergency Core Cooling System (ECCS) piping and spargers), nuclear fuel, and the control rods and control rod drive system (including hydraulic supply and accumulators). Special tools required for reactor system maintenance, inspection and testing are also covered.

  12. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  13. Polished rod liner puller assmbly

    SciTech Connect

    Baxter, B.V.

    1990-01-02

    This patent describes a polished rod liner puller assembly operable with a well casing head assembly to remove a polished rod liner member of a polished rod liner assembly from a well. It comprises: a work table assembly operable to be placed around the well casing head assembly and enclose the polished rod liner assembly; a base plate assembly mounted on the work table assembly; a piston and cylinder jack assembly mounted on the base plate assembly and extended upwardly therefrom; and a winged rod clamp assembly connectable to the piston and cylinder jack assembly and to a polished rod member of the polished rod liner assembly and operable on actuation of the piston and cylinder jack assembly to axially move the polished rod member and the polished rod liner member to remove the polished rod liner member from the well.

  14. Nuclear reactor safety device

    DOEpatents

    Hutter, Ernest

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  15. Reactor physics analyses of the advanced neutron source three-element core

    SciTech Connect

    Gehin, J.C.

    1995-08-01

    A reactor physics analysis was performed for the Advanced Neutron Source reactor with a three-element core configuration. The analysis was performed with a two-dimensional r-z 20-energy-group finite-difference diffusion theory model of the 17-d fuel cycle. The model included equivalent r-z geometry representations of the central control rods, the irradiation and production targets, and reflector components. Calculated quantities include fuel cycle parameters, fuel element power distributions, unperturbed neutron fluxes in the reflector and target regions, reactivity perturbations, and neutron kinetics parameters.

  16. Nonlinearity and noise at the rod - rod bipolar cell synapse

    PubMed Central

    Trexler, E. Brady; Casti, Alexander R.R.; Zhang, Yu

    2010-01-01

    In the retina, rod bipolar (RBP) cells synapse with many rods, and suppression of rod outer segment and synaptic noise is necessary for their detection of rod single photon responses (SPRs). Depending on the rods’ signal to noise ratio (SNR), the suppression mechanism will likely eliminate some SPRs as well, resulting in decreased quantum efficiency. We examined this synapse in rabbit, where 100 rods converge onto each RBP. Suction electrode recordings showed that rabbit rod SPRs were difficult to distinguish from noise (independent SNR estimates were 2.3 and 2.8). Nonlinear transmission from rods to RBPs improved response detection (SNR = 8.7), but a large portion of the rod SPRs were discarded. For the dimmest flashes, the loss approached 90%. Despite the high rejection ratio, noise of two distinct types were apparent in the RBP traces: low amplitude rumblings and discrete events that resembled the SPR. The SPR-like event frequency suggests they result from thermal isomerizations of rhodopsin, which occured at the rate 0.033 s−1rod−1. The presence of low amplitude noise is explained by a sigmoidal input-output relationship at the rod - RBP synapse and the input of noisy rods. The rabbit rod SNR and RBP quantum efficiency are the lowest yet reported, suggesting that the quantum efficiency of the rod - RBP synapse may depend on the SNR in rods. These results point to the possibility that fewer photoisomerizations are discarded for species such as primate, which has a higher rod SNR. PMID:21047445

  17. Verification of maximum radial power peaking factor due to insertion of FPM-LEU target in the core of RSG-GAS reactor

    SciTech Connect

    Setyawan, Daddy; Rohman, Budi

    2014-09-30

    Verification of Maximum Radial Power Peaking Factor due to insertion of FPM-LEU target in the core of RSG-GAS Reactor. Radial Power Peaking Factor in RSG-GAS Reactor is a very important parameter for the safety of RSG-GAS reactor during operation. Data of radial power peaking factor due to the insertion of Fission Product Molybdenum with Low Enriched Uranium (FPM-LEU) was reported by PRSG to BAPETEN through the Safety Analysis Report RSG-GAS for FPM-LEU target irradiation. In order to support the evaluation of the Safety Analysis Report incorporated in the submission, the assessment unit of BAPETEN is carrying out independent assessment in order to verify safety related parameters in the SAR including neutronic aspect. The work includes verification to the maximum radial power peaking factor change due to the insertion of FPM-LEU target in RSG-GAS Reactor by computational method using MCNP5and ORIGEN2. From the results of calculations, the new maximum value of the radial power peaking factor due to the insertion of FPM-LEU target is 1.27. The results of calculations in this study showed a smaller value than 1.4 the limit allowed in the SAR.

  18. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  19. PBF Reactor Building (PER620). Camera looks into reactor vessel. Control ...

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

    PBF Reactor Building (PER-620). Camera looks into reactor vessel. Control rods are positioned at outer perimeter; transient rods, at inner perimeter. Photographer: Larry Page. Date: November 2, 1972. INEEL negative no. 72-5266 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  20. 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.

  1. 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.

  2. Internal Control Rod Drive Mechanisms, Design Options for IRIS

    SciTech Connect

    Conway, Lawrence E.; Petrovic, Bojan

    2004-07-01

    IRIS (International Reactor Innovative and Secure) is a medium-power (335 MWe) PWR with an integral, primary circuit configuration, where all the reactor coolant system components are contained within the reactor vessel. This integral configuration is a key reason for the success of IRIS' 'safety-by-design' approach, whereby accident initiators are eliminated or the accident consequences and/or frequency are reduced. The most obvious example of the IRIS safety by design approach is the elimination of large LOCA's, since the integral reactor coolant system has no large loop piping. Another serious accident scenario that is being addressed in IRIS is the postulated ejection of a reactor control cluster assembly (RCCA). This accident initiator can be eliminated by locating the RCCA drive mechanisms (CRDMs) inside the reactor vessel. This eliminates the mechanical drive rod penetration between the RCCA and the external CRDM, eliminating the potential for differential pressure across the pressure boundary, and thus eliminating 'by design' the possibility for rod ejection accident. Moreover, the elimination of the 'large' drive-rod penetrations and the external CRDM pressure housings decreases the likelihood of boric acid leakage and subsequent corrosion of the reactor pressure boundary (like the Davis-Besse incident). This paper will discuss the IRIS top level design requirements and objectives for internal CRDMs, and provide examples candidate designs and their specific performance characteristics. (authors)

  3. 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.

  4. 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.

  5. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  6. Welded oil well pump rod

    SciTech Connect

    Hughes, R.F.

    1986-06-10

    A friction welded multiple component oil well sucker rod is described which consists of an elongated cylindrical rod section and apposed coupling end portions welded to opposite ends of the rod section, the coupling end portions being of a nominal maximum diameter at least 1.5 times greater than the rod section and including means for connecting the sucker rod to an adjacent rod in end to end relationship. The couplings end portions each include an axial tapered portion between the connecting means and an end face adapted to be butted to the rod section, the coupling end portions being butted against the opposed end portions of the rod section during a friction welding operation to form a radially outward projecting bulge of displaced material on the rod section and the coupling end portions, respectively. A greater cross-sectional area is formed at the transition of the rod section to the coupling end portion to reduce the unit tensile stress on the sucker rod in the vicinity of the weld, wherein the displaced material is machined to form a tapered surface between the rod section and the axial tapered portion of the coupling end portion, the tapered surface having an angle of taper with respect to the longitudinal axis of the sucker rod less than the angle of taper of the coupling end portion.

  7. 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.

  8. Class D sucker rods

    SciTech Connect

    Woodings, R. T.

    1984-10-23

    It has been found that API Class D sucker rods can be made inexpensively from low-alloy, low-cost steel by following a suitable induction-normalizing process and using a suitable steel to which there has been added 0.07 to 0.15 percent of vanadium.

  9. An RF-powered micro-reactor for the detection of astrobiological target molecules on planetary bodies

    NASA Astrophysics Data System (ADS)

    Scott, Valerie J.; Tse, Margaret; Shearn, Michael J.; Siegel, Peter H.; Amashukeli, Xenia

    2012-08-01

    We describe a sample-processing micro-reactor that utilizes 60 GHz RF radiation with approximately 730 mW of output power. The instrument design and performance characterization are described and then illustrated with modeling and experimental studies. The micro-reactor's efficiency on affecting hydrolysis of chemical bonds similar to those within large complex molecules was demonstrated: a disaccharide—sucrose—was hydrolyzed completely under micro-reactor conditions. The products of the micro-reactor-facilitated hydrolysis were analyzed using mass spectroscopy and proton nuclear magnetic resonance analytical techniques.

  10. An RF-powered micro-reactor for the detection of astrobiological target molecules on planetary bodies.

    PubMed

    Scott, Valerie J; Tse, Margaret; Shearn, Michael J; Siegel, Peter H; Amashukeli, Xenia

    2012-08-01

    We describe a sample-processing micro-reactor that utilizes 60 GHz RF radiation with approximately 730 mW of output power. The instrument design and performance characterization are described and then illustrated with modeling and experimental studies. The micro-reactor's efficiency on affecting hydrolysis of chemical bonds similar to those within large complex molecules was demonstrated: a disaccharide-sucrose-was hydrolyzed completely under micro-reactor conditions. The products of the micro-reactor-facilitated hydrolysis were analyzed using mass spectroscopy and proton nuclear magnetic resonance analytical techniques.

  11. BWR control-rod cobalt-alloy replacement. Final report

    SciTech Connect

    Aldred, P.

    1982-03-01

    Cobalt base pin and roller alloys in BWR Control Rods are a source for the Co-60 isotope which contributes to radiation buildup in the BWR core, the recirculation piping system and the spent fuel pool. It thereby influences personnel radiation exposure during BWR plant maintenance. The program objectives were (a) to identify non-cobalt alloys which could potentially replace the cobalt alloys, (b) evaluate the alloys by testing to qualify them for in-reactor surveillance testing, and (c) to initiate reactor tests at 2 BWRs. Wear resistance, an important requirement for pins and rollers, was measured in a simulated BWR environment (excluding irradiation). Prototypic wear tests were emphasized and a prototype control rod drive test facility was used to evaluate several pin and roller alloy combinations during prototype control rod operations.

  12. Physics analysis of the gang partial rod drive event

    SciTech Connect

    Boman, C.; Frost, R.L.

    1992-08-01

    During the routine positioning of partial-length control rods in Gang 3 on the afternoon of Monday, July 27, 1992, the partial-length rods continued to drive into the reactor even after the operator released the controlling toggle switch. In response to this occurrence, the Safety Analysis and Engineering Services Group (SAEG) requested that the Applied Physics Group (APG) analyze the gang partial rod drive event. Although similar accident scenarios were considered in analysis for Chapter 15 of the Safety Analysis Report (SAR), APG and SAEG conferred and agreed that this particular type of gang partial-length rod motion event was not included in the SAR. This report details this analysis.

  13. Assessment of precision gamma scanning for inspecting LWR fuel rods. Final report

    SciTech Connect

    Phillips, J.R.; Barnes, B.K.; Barnes, M.L.; Hamlin, D.K.; Medina-Ortega, E.G.

    1981-07-01

    Reconstruction of the radial two-dimensional distributions of fission products using projections obtained by nondestructive gamma scanning was evaluated. The filtered backprojection algorithm provided the best reconstruction for simulated gamma-ray sources, as well as for actual irradiated fuel material. Both a low-burnup (11.5 GWd/tU) light-water reactor fuel rod and a high-burnup (179.1 GWd/tU) fast breeder reactor fuel rod were examined using this technique.

  14. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  15. CONTROL ROD ROTATING MECHANISM

    DOEpatents

    Baumgarten, A.; Karalis, A.J.

    1961-11-28

    A threaded rotatable shaft is provided which rotates in response to linear movement of a nut, the shaft being surrounded by a pair of bellows members connected to either side of the nut to effectively seal the reactor from leakage and also to store up energy to shut down the reactor in the event of a power failure. (AEC)

  16. Computer program for optimal BWR congtrol rod programming

    SciTech Connect

    Taner, M.S.; Levine, S.H.; Carmody, J.M.

    1995-12-31

    A fully automated computer program has been developed for designing optimal control rod (CR) patterns for boiling water reactors (BWRs). The new program, called OCTOPUS-3, is based on the OCTOPUS code and employs SIMULATE-3 (Ref. 2) for the analysis. There are three aspects of OCTOPUS-3 that make it successful for use at PECO Energy. It incorporates a new feasibility algorithm that makes the CR design meet all constraints, it has been coupled to a Bourne Shell program 3 to allow the user to run the code interactively without the need for a manual, and it develops a low axial peak to extend the cycle. For PECO Energy Co.`s limericks it increased the energy output by 1 to 2% over the traditional PECO Energy design. The objective of the optimization in OCTOPUS-3 is to approximate a very low axial peaked target power distribution while maintaining criticality, keeping the nodal and assembly peaks below the allowed maximum, and meeting the other constraints. The user-specified input for each exposure point includes: CR groups allowed-to-move, target k{sub eff}, and amount of core flow. The OCTOPUS-3 code uses the CR pattern from the previous step as the initial guess unless indicated otherwise.

  17. Semiconductor Quantum Rods as Single Molecule FluorescentBiological Labels

    SciTech Connect

    Fu, Aihua; Gu, Weiwei; Boussert, Benjamine; Koski, Kristie; Gerion, Daniele; Manna, Liberato; Le Gros, Mark; Larabell, Carolyn; Alivisatos, A. Paul

    2006-05-29

    In recent years, semiconductor quantum dots have beenapplied with great advantage in a wide range of biological imagingapplications. The continuing developments in the synthesis of nanoscalematerials and specifically in the area of colloidal semiconductornanocrystals have created an opportunity to generate a next generation ofbiological labels with complementary or in some cases enhanced propertiescompared to colloidal quantum dots. In this paper, we report thedevelopment of rod shaped semiconductor nanocrystals (quantum rods) asnew fluorescent biological labels. We have engineered biocompatiblequantum rods by surface silanization and have applied them fornon-specific cell tracking as well as specific cellular targeting. Theproperties of quantum rods as demonstrated here are enhanced sensitivityand greater resistance for degradation as compared to quantum dots.Quantum rods have many potential applications as biological labels insituations where their properties offer advantages over quantumdots.

  18. Sucker rod centralizer

    SciTech Connect

    Rivas, O.; Newski, A.

    1989-10-03

    This patent describes a device for centralizing at least one sucker rod within a production pipe downhole in a well and for reducing frictional forces between the pipe and at least one sucker rod. It comprises an elongate, substantially cylindrical body member having a longitudinal axis, a plurality of slots within the member and a rotatable member mounted within each slot, each of the plurality of slots has its major dimension along a first axis parallel to the longitudinal axis of the body member and is oriented with respect to the other seats so as to form a helicoidal array for maximizing the total surface contact area between the rotatable members and the pipe and for decreasing the forces acting on each rotatable member.

  19. Experimental plan and design of two experiments for graphite irradiation at temperatures up to 1500 °C in the target region of the high flux isotope reactor

    NASA Astrophysics Data System (ADS)

    McDuffee, J. L.; Burchell, T. D.; Heatherly, D. W.; Thoms, K. R.

    2008-10-01

    Two irradiation capsules have been designed for the target region of the high flux isotope reactor (HFIR). The objective is to provide dimensional change and physical property data for four candidate next generation nuclear plant (NGNP) graphites. The capsules will reach peak doses of ˜1.59 and ˜4.76 dpa, respectively, at temperatures of 900, 1200, and 1500 °C.

  20. Mechanism for Selective Synaptic Wiring of Rod Photoreceptors into the Retinal Circuitry and Its Role in Vision.

    PubMed

    Cao, Yan; Sarria, Ignacio; Fehlhaber, Katherine E; Kamasawa, Naomi; Orlandi, Cesare; James, Kiely N; Hazen, Jennifer L; Gardner, Matthew R; Farzan, Michael; Lee, Amy; Baker, Sheila; Baldwin, Kristin; Sampath, Alapakkam P; Martemyanov, Kirill A

    2015-09-23

    In the retina, rod and cone photoreceptors form distinct connections with different classes of downstream bipolar cells. However, the molecular mechanisms responsible for their selective connectivity are unknown. Here we identify a cell-adhesion protein, ELFN1, to be essential for the formation of synapses between rods and rod ON-bipolar cells in the primary rod pathway. ELFN1 is expressed selectively in rods where it is targeted to the axonal terminals by the synaptic release machinery. At the synapse, ELFN1 binds in trans to mGluR6, the postsynaptic receptor on rod ON-bipolar cells. Elimination of ELFN1 in mice prevents the formation of synaptic contacts involving rods, but not cones, allowing a dissection of the contributions of primary and secondary rod pathways to retinal circuit function and vision. We conclude that ELFN1 is necessary for the selective wiring of rods into the primary rod pathway and is required for high sensitivity of vision.

  1. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Untermyer, S.; Hutter, E.

    1959-08-01

    This patent relates to "shadow" control of a nuclear reactor. The control means comprises a plurality ot elongated rods disposed adjacent and parallel to each other, The morphology and effects of gases generated within sections of neutron absorbing materials and equal length sections of neutron permeable materials together with means for longitudinally pcsitioning the rcds relative to each other.

  2. Sucker rod guide

    SciTech Connect

    White, R.C.

    1988-10-25

    This patent describes an improved guide for use in a string of sucker rods for reciprocation in a tubing string in a borehole, the sucker rods having threaded male ends, the guide comprising: an elongated upright cylindrical member of external diameter less than the internal diameter of tubing in which it is to be used, the member having sucker rod receiving female threaded openings at the upper and lower ends, the threaded openings being coaxial of the member cylindrical axis whereby the member may be positioned in a string of sucker rods, and including a plurality of spaced-apart parallel sided slots within the member, each slot being of semi-circular configuration and of depth greater than the radius and less than the diameter of the cylindrical member, the sidewalls of each slot being parallel to and equally spaced from a plane of the member cylindrical axis; the member having an axle bore therein for each of the slots, the axle bores being parallel and spaced apart from each other, a plane of the axis of each bore being perpendicular the member cylindrical axis and the axis of each bore being displaced away from the member cylindrical axis; an axle received in each axle bore; and a wheel received on each axle the diameter of each wheel being approximately the diameter of the cylindrical member, the periphery of each wheel extending beyond the member cylindrical wall whereby the wheels are positioned to engage and roll on the internal cylindrical surface of tubing, the planes of adjacent slots in the member being rotationally displaced from each other, a portion of each wheel extending beyond the cylindrical surface of the member, the opposed portion of each wheel being within the confines of the member cylindrical surface whereby each wheel can contact a tubing wall at only one point on its cylindrical surface.

  3. 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)

  4. Sucker rod coupling

    SciTech Connect

    Klyne, A.A.

    1986-11-11

    An anti-friction sucker rod coupling is described for connecting a pair of sucker rods and centralizing them in a tubing string, comprising: an elongate, rigid, substantially cylindrical body member, each end of the body member forming means for threadably connecting the body member with a sucker rod. The body member further forms a transversely extending, substantially diametric, generally vertical slot extending therethrough. The body member further forms a pin bore, such pin bore extending transversely through the body member so as to intersect the slot substantially perpendicularly; a wheel member positioned within the slot to rotate in a generally vertical plane. The wheel member has a portion thereof extending beyond the periphery of the body member to engage the inner surface of the tubing string and centralize the coupling; and a pin mounted in the pin bore and supporting member thereon, whereby the wheel member is rotatable within the slot; the wheel member having sufficient clearance between its side surfaces and the wall surfaces of the slot, when the wheel member is centered in the slot on the pin, whereby the wheel member may shift along the pin to assist in ejecting sand and oil from the slot.

  5. Rod locking device

    SciTech Connect

    Troxell, J.N. Jr.

    1986-07-22

    A ram locking apparatus used on a blowout preventer is described having a housing, a ram, ram actuating means having a closing side and a retracted side and a tail rod having its inner end connected to the ram actuating means and its outer end engaged by the apparatus to lock the ram. The apparatus consists of: a lock housing having a closed end and a hollow interior connected to the exterior of the preventer housing in which the tail rod is positioned, a body positioned within the lock housing, a primary piston, a lost motion connection between the primary piston and the body, a lock piston associated with the primary piston and movable axially with respect to the primary piston, a tapered split locking ring interconnected to the lock piston, wedging means with the split locking ring, and means for supplying fluid under pressure into the lock housing for movement of the pistons, the initial pressure on the primary pistons causing movement of the body to engage the ram tail rod and subsequently moving the lock piston relative to the wedging means and to thereby wedge the split locking ring against the interior of the lock housing to lock the body therein against movement in the lock housing.

  6. 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.

  7. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    SciTech Connect

    Clayton, J C

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated.

  8. Influence of control rod worth interactions on LMFBR control systems design

    SciTech Connect

    Lake, J. A.; Rittenberger, R. V.; Rathbun, R. W.

    1980-01-01

    These design criteria are interpreted to define the reactivity worth requirements for the primary and secondary control systems in terms of the minimum control systems capability under faulted conditions which will assure that the reactor can be safely shut down. The faulted conditions are postulated to occur upon the simultaneous failure of one of the redundant safety control systems to scram, a stuck rod in the scramming system, and a reactivity insertion resulting from the uncontrolled withdrawal of the highest worth control rod inserted in the reactor. The resulting positive reactivity insertion from the rod runout envelopes other postulated operational faults and is imposed on the shutdown requirements of both the primary and secondary control systems. In order to determine the minimum shutdown capability, an evaluation is made of the worst combination of control rod runout (reactivity fault) and stuck rod worth.

  9. An Investigation into the Transportation of Irradiated Uranium/Aluminum Targets from a Foreign Nuclear Reactor to the Chalk River Laboratories Site in Ontario, Canada - 12249

    SciTech Connect

    Clough, Malcolm; Jackson, Austin

    2012-07-01

    This investigation required the selection of a suitable cask and development of a device to hold and transport irradiated targets from a foreign nuclear reactor to the Chalk River Laboratories in Ontario, Canada. The main challenge was to design and validate a target holder to protect the irradiated HEU-Al target pencils during transit. Each of the targets was estimated to have an initial decay heat of 118 W prior to transit. As the targets have little thermal mass the potential for high temperature damage and possibly melting was high. Thus, the primary design objective was to conceive a target holder to dissipate heat from the targets. Other design requirements included securing the targets during transportation and providing a simple means to load and unload the targets while submerged five metres under water. A unique target holder (patent pending) was designed and manufactured together with special purpose experimental apparatus including a representative cask. Aluminum dummy targets were fabricated to accept cartridge heaters, to simulate decay heat. Thermocouples were used to measure the temperature of the test targets and selected areas within the target holder and test cask. After obtaining test results, calculations were performed to compensate for differences between experimental and real life conditions. Taking compensation into consideration the maximum target temperature reached was 231 deg. C which was below the designated maximum of 250 deg. C. The design of the aluminum target holder also allowed generous clearance to insert and unload the targets. This clearance was designed to close up as the target holder is placed into the cavity of the transport cask. Springs served to retain and restrain the targets from movement during transportation as well as to facilitate conductive heat transfer. The target holder met the design requirements and as such provided data supporting the feasibility of transporting targets over a relatively long period of time

  10. Nuclear reactor reflector

    DOEpatents

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  11. Nuclear reactor reflector

    DOEpatents

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  12. REACTOR CONTROL SYSTEM

    DOEpatents

    MacNeill, J.H.; Estabrook, J.Y.

    1960-05-10

    A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

  13. COOLED NEUTRONIC REACTOR

    DOEpatents

    Binner, C.R.; Wilkie, C.B.

    1958-03-18

    This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

  14. 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.

  15. 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.

  16. Cone rod dystrophies.

    PubMed

    Hamel, Christian P

    2007-01-01

    Cone rod dystrophies (CRDs) (prevalence 1/40,000) are inherited retinal dystrophies that belong to the group of pigmentary retinopathies. CRDs are characterized by retinal pigment deposits visible on fundus examination, predominantly localized to the macular region. In contrast to typical retinitis pigmentosa (RP), also called the rod cone dystrophies (RCDs) resulting from the primary loss in rod photoreceptors and later followed by the secondary loss in cone photoreceptors, CRDs reflect the opposite sequence of events. CRD is characterized by primary cone involvement, or, sometimes, by concomitant loss of both cones and rods that explains the predominant symptoms of CRDs: decreased visual acuity, color vision defects, photoaversion and decreased sensitivity in the central visual field, later followed by progressive loss in peripheral vision and night blindness. The clinical course of CRDs is generally more severe and rapid than that of RCDs, leading to earlier legal blindness and disability. At end stage, however, CRDs do not differ from RCDs. CRDs are most frequently non syndromic, but they may also be part of several syndromes, such as Bardet Biedl syndrome and Spinocerebellar Ataxia Type 7 (SCA7). Non syndromic CRDs are genetically heterogeneous (ten cloned genes and three loci have been identified so far). The four major causative genes involved in the pathogenesis of CRDs are ABCA4 (which causes Stargardt disease and also 30 to 60% of autosomal recessive CRDs), CRX and GUCY2D (which are responsible for many reported cases of autosomal dominant CRDs), and RPGR (which causes about 2/3 of X-linked RP and also an undetermined percentage of X-linked CRDs). It is likely that highly deleterious mutations in genes that otherwise cause RP or macular dystrophy may also lead to CRDs. The diagnosis of CRDs is based on clinical history, fundus examination and electroretinogram. Molecular diagnosis can be made for some genes, genetic counseling is always advised. Currently

  17. Additional information for impact response of the restart safety rods

    SciTech Connect

    Yau, W.W.F.

    1991-10-14

    WSRC-RP-91-677 studied the structural response of the safety rods under the conditions of brake failure and accidental release. It was concluded that the maximum impact loading to the safety rod is 6020 pounds based on conservative considerations that energy dissipation attributable to fluid resistance and reactor superstructure flexibility. The staffers of the Defense Nuclear Facility Safety Board reviewed the results and inquired about the extent of conservatism. By request of the RESTART team, I reassessed the impact force due to these conservative assumptions. This memorandum reports these assessments.

  18. Glassy materials investigated for nuclear reactor applications

    NASA Technical Reports Server (NTRS)

    Lynch, E. D.

    1968-01-01

    Studies determine the feasibility of preparing fuel-bearing glasses and glasses bearing neutron-absorbing materials for use as crystalline fuel and control rods for reactors. Properties investigated were devitrification resistance, urania solubility, and density.

  19. Conversion of Molybdenum-99 production process to low enriched uranium: Neutronic and thermal hydraulic analyses of HEU and LEU target plates for irradiation in Pakistan Research Reactor-1

    NASA Astrophysics Data System (ADS)

    Mushtaq, Ahmad; Iqbal, Masood; Bokhari, Ishtiaq Hussain; Mahmood, Tayyab; Muhammad, Atta

    2012-09-01

    Technetium-99m, the daughter product of Molybdenum-99 is the most widely needed radionuclide for diagnostic studies in Pakistan. Molybdenum-99 Production Facility has been established at PINSTECH. Highly enriched uranium (93% 235U) U/Al alloy targets have been irradiated in Pakistan Research Reactor-1 (PARR-1) for the generation of fission Mo-99, while basic dissolution technique is used for separation of Mo-99 from target matrix activity. In line with the international objective of minimizing and eventually eliminating the use of HEU in civil commerce, national and international efforts have been underway to shift the production of medical isotopes from HEU to LEU (LEU; <20% 235U enrichment) targets. To achieve the equivalent amount of 99Mo with LEU targets, approximately 5 times uranium is needed. LEU aluminum uranium dispersion target has been developed, which may replace existing HEU aluminum/uranium alloy targets for production of 99Mo using basic dissolution technique. Neutronic and thermal hydraulic calculations were performed for safe irradiation of targets in the core of PARR-1.

  20. 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.

  1. Sucker rod assembly and method

    SciTech Connect

    Pagan, A.J.

    1986-07-01

    An improved sucker rod assembly is described comprising, in combination: a. a sucker rod; and b. a pair of fittings secured to opposite ends of the rod, each fitting including: i. a rigid elongated casing having interior surfaces defining an open front end and cavity extending rearwardly from the open front end in which cavity one end of the sucker rod is disposed, the side portions of the interior surfaces being contoured to define, with the side portions of the sucker rod end a single, annular elongated tapered wedge-shaped space; and ii. anchoring means filling the space and bonding to the side portions of the rod end to lock the rod end in place, the anchoring means having a narrower diameter at the front end thereof than at about the rear end thereof and being generally frusto-conical, the anchoring means comprising a plurality of separate rigid inserts, the interior surfaces of which collectively define a central elongated passageway in which the rod end is received, the interior surfaces of the inserts being tightly bonded to the side portions of the rod, and the inserts being bonded to each other along the contact lines therebetween to form a unitary structure.

  2. Implementation of CTRLPOS, a VENTURE module for control rod position criticality searches, control rod worth curve calculations, and general criticality searches

    SciTech Connect

    Smith, L.A.; Renier, J.P.

    1994-06-01

    A module in the VENTURE reactor analysis code system, CTRLPOS, is developed to position control rods and perform control rod position criticality searches. The module is variably dimensioned so that calculations can be performed with any number of control rod banks each having any number of control rods. CTRLPOS can also calculate control rod worth curves for a single control rod or a bank of control rods. Control rod depletion can be calculated to provide radiation source terms. These radiation source terms can be used to predict radiation doses to personnel and estimate the shielding and long-term storage requirements for spent control rods. All of these operations are completely automated. The numerous features of the module are discussed in detail. The necessary input data for the CTRLPOS module is explained. Several sample problems are presented to show the flexibility of the module. The results presented with the sample problems show that the CTRLPOS module is a powerful tool which allows a wide variety of calculations to be easily performed.

  3. REACTOR CONTROL

    DOEpatents

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  4. PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

    DOEpatents

    Puechl, K.H.

    1963-09-24

    A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

  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. Analysis of dose rates received around the storage pool for irradiated control rods in a BWR nuclear power plant.

    PubMed

    Ródenas, J; Abarca, A; Gallardo, S

    2011-08-01

    BWR control rods are activated by neutron reactions in the reactor. The dose produced by this activity can affect workers in the area surrounding the storage pool, where activated rods are stored. Monte Carlo (MC) models for neutron activation and dose assessment around the storage pool have been developed and validated. In this work, the MC models are applied to verify the expected reduction of dose when the irradiated control rod is hanged in an inverted position into the pool.

  7. 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.

  8. 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.

  9. Sucker rod pump

    SciTech Connect

    Brewer, J.R.

    1992-04-14

    This patent describes a subsurface well pump, it comprises: a working barrel; a plunger which reciprocates along the vertical axis within the working barrel between an upper and lower position; a rod connected to the plunger and extending to a means for providing reciprocating force; a well string extending from the top of the working barrel to the surface; an outlet check valve which permits flow to exit the working barrel into the well string and does not permit flow to exit the well string into the working barrel; and an inlet check valve which permits flow into the working barrel from outside of the subsurface pump, the inlet check valve being above the top position of the plunger, the inlet check valve having a cross sectional flow area about equal to or greater than the horizontal cross sectional area of the working barrel, and the inlet check valve being a hinged flapper valve.

  10. Reactivity control assembly for nuclear reactor. [LMFBR

    DOEpatents

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  11. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  12. Scalable Methods for Uncertainty Quantification, Data Assimilation and Target Accuracy Assessment for Multi-Physics Advanced Simulation of Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Khuwaileh, Bassam

    High fidelity simulation of nuclear reactors entails large scale applications characterized with high dimensionality and tremendous complexity where various physics models are integrated in the form of coupled models (e.g. neutronic with thermal-hydraulic feedback). Each of the coupled modules represents a high fidelity formulation of the first principles governing the physics of interest. Therefore, new developments in high fidelity multi-physics simulation and the corresponding sensitivity/uncertainty quantification analysis are paramount to the development and competitiveness of reactors achieved through enhanced understanding of the design and safety margins. Accordingly, this dissertation introduces efficient and scalable algorithms for performing efficient Uncertainty Quantification (UQ), Data Assimilation (DA) and Target Accuracy Assessment (TAA) for large scale, multi-physics reactor design and safety problems. This dissertation builds upon previous efforts for adaptive core simulation and reduced order modeling algorithms and extends these efforts towards coupled multi-physics models with feedback. The core idea is to recast the reactor physics analysis in terms of reduced order models. This can be achieved via identifying the important/influential degrees of freedom (DoF) via the subspace analysis, such that the required analysis can be recast by considering the important DoF only. In this dissertation, efficient algorithms for lower dimensional subspace construction have been developed for single physics and multi-physics applications with feedback. Then the reduced subspace is used to solve realistic, large scale forward (UQ) and inverse problems (DA and TAA). Once the elite set of DoF is determined, the uncertainty/sensitivity/target accuracy assessment and data assimilation analysis can be performed accurately and efficiently for large scale, high dimensional multi-physics nuclear engineering applications. Hence, in this work a Karhunen-Loeve (KL

  13. Safety rod/thimble melt failure characterization experiments

    SciTech Connect

    Stoots, C.M.; Hawkes, G.L.

    1992-05-01

    The Department of Energy (DOE) requested that he INEL perform experiments to study the thermal failure characteristics of a simulated Savannah River Site nuclear reactor safety rod and its surrounding thimble assembly. An electrically heated stainless steel rod simulated a reactor safety rod located eccentrically or concentrically within a perforated aluminum guide tube or thimble. A total of 37 experiments were conducted for a range of power levels and safety rod/thimble relative orientations. Video tapes were made of the four failure tests that were conducted to the melting point of the thimble. Although the primary emphasis of the experiments were to characterize the melting of the thimble qualitatively, experimental transient measurements included heater voltage and current, heater surface temperatures, aluminum thimble temperatures, and ambient temperature. Numerical studies were also performed in support of the experiments and data interpretation. Two finite element models were created to model the heat conduction-radiation between the stainless steel heater and thimble. The predicted temperatures were in good agreement with the experimental results.

  14. 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

  15. Eulerian formulation of elastic rods

    NASA Astrophysics Data System (ADS)

    Huynen, Alexandre; Detournay, Emmanuel; Denoël, Vincent

    2016-06-01

    In numerous biological, medical and engineering applications, elastic rods are constrained to deform inside or around tube-like surfaces. To solve efficiently this class of problems, the equations governing the deflection of elastic rods are reformulated within the Eulerian framework of this generic tubular constraint defined as a perfectly stiff normal ringed surface. This reformulation hinges on describing the rod-deformed configuration by means of its relative position with respect to a reference curve, defined as the axis or spine curve of the constraint, and on restating the rod local equilibrium in terms of the curvilinear coordinate parametrizing this curve. Associated with a segmentation strategy, which partitions the global problem into a sequence of rod segments either in continuous contact with the constraint or free of contact (except for their extremities), this re-parametrization not only trivializes the detection of new contacts but also transforms these free boundary problems into classic two-points boundary-value problems and suppresses the isoperimetric constraints resulting from the imposition of the rod position at the extremities of each rod segment.

  16. 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.

  17. Rod coupling for oil well sucker rods and the like

    SciTech Connect

    Bowers, R.

    1986-07-29

    A coupling is described for joining solid reciprocating sucker rods to form a rod string in a well pump or the like comprising a unitary metal sleeve having an axial threaded bore and an irregular outer surface, and a homogeneous and non-fibrous coating on the sleeve over the outer surface providing an externally substantially cylindrical coupling, the coating comprising a flexible and abrasive resistant thermoplastic hydrourethane polymer formed on the irregular outer surface of the sleeve while in the molten state.

  18. Examination of cadmium safety rod thermal test specimens and failure mechanism evaluation

    SciTech Connect

    Thomas, J.K.; Peacock, H.B.; Iyer, N.C.

    1992-01-01

    The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of a hypothetical LOCA event. Accordingly, an experimental cadmium safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. Companion reports describe the experiments and a structural evaluation (finite element analysis) of the safety rod. This report deals primarily with the examination of the test specimens, evaluation of possible failure mechanisms, and confirmatory separate effects experiments. It is concluded that the failures observed in the cadmium safety rod thermal tests which occurred at low temperature (T < 600{degrees}C) with slow thermal ramp rates (slow cladding strain rates) resulted from localized dissolution of the stainless steel cladding by the cadmium/aluminum solution and subsequent ductility exhaustion and rupture. The slow thermal ramp rate is believed to be the root cause for the failures; specifically, the slow ramp rate led to localized cladding shear deformation which ruptured the protective oxide film on the cladding inner surface and allowed dissolution to initiate. The test results and proposed failure mechanism support the conclusion that the rods would not fail below 500{degrees}C even at slow ramp rates. The safety rod thermal test specimen failures which occurred at high temperature (T > 800{degrees}C) with fast thermal ramp rates are concluded to be mechanical in nature without significant environmental degradation. Based on these tests, tasks were initiated to design and manufacture B{sub 4}C safety rods to replace the cadmium safety rods. The B{sub 4}C safety rods have been manufactured at this time and it is currently planned to charge them to the reactor in the near future. 60 refs.

  19. Rod coupling with mounted guide

    SciTech Connect

    Bair, M.L.

    1987-05-26

    This patent describes a well sucker rod string, in a well bore, the combination comprising: an axially elongated coupling section having threads at axially opposite ends thereof for coupling to and between successive sucker rods in the rod string, to transmit string loading. The section has first and second exposed surfaces adjacent an end of the section, and a third surface located between the first and second exposed surfaces; a rod guide consisting of molded plastic material extending about and bonded to the section third surface to project outwardly therefrom for engagement with the well bore during up and down stroking of the string; and one annular groove sunk in the section between the first and third surfaces, and another annular groove sunk in the section between the second and third surfaces. The depth of the one groove is less than about 15% of the radius of the section at the first surface.

  20. THERMAL NUCLEAR REACTOR

    DOEpatents

    Fenning, F.W.; Jackson, R.F.

    1957-09-24

    Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.

  1. TRAC analysis of design basis events for the accelerator production of tritium target/blanket

    SciTech Connect

    Lin, J.C.; Elson, J.

    1997-08-01

    A two-loop primary cooling system with a residual heat removal system was designed to mitigate the heat generated in the tungsten neutron source rods inside the rungs of the ladders and the shell of the rungs. The Transient Reactor Analysis Code (TRAC) was used to analyze the thermal-hydraulic behavior of the primary cooling system during a pump coastdown transient; a cold-leg, large-break loss-of-coolant accident (LBLOCA); a hot-leg LBLOCA; and a target downcomer LBLOCA. The TRAC analysis results showed that the heat generated in the tungsten neutron source rods can be mitigated by the primary cooling system for the pump coastdown transient and all the LBLOCAs except the target downcomer LBLOCA. For the target downcomer LBLOCA, a cavity flood system is required to fill the cavity with water at a level above the large fixed headers.

  2. 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

  3. Tensile and burst tests in support of the cadmium safety rod failure evaluation

    SciTech Connect

    Thomas, J.K.

    1992-02-01

    The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This report discusses confirmatory separate effects tests conducted to support the evaluation of failures observed in the safety rod thermal tests. As part of the failure evaluation, the potential for liquid metal embrittlement (LME) of the safety rod cladding by cadmium (Cd) -- aluminum (Al) solutions was examined. Based on the test conditions, literature data, and U-Bend tests, its was concluded that the SS304 safety rod cladding would not be subject to LME by liquid Cd-Al solutions under conditions relevant to the safety rod thermal tests or gamma heating accident. To confirm this conclusion, tensile tests on SS304 specimens were performed in both air and liquid Cd-Al solutions with the range of strain rates, temperatures, and loading conditions spanning the range relevant to the safety rod thermal tests and gamma heating accident.

  4. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  5. 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.

  6. RP cone-rod degeneration.

    PubMed Central

    Heckenlively, J R

    1987-01-01

    A group of patients with progressive retinal degeneration and visual field loss, who meet the basic definition of RP were investigated to better define the relationship of the findings on the ERG with clinical characteristics such as visual field size, presence or absence of scotomata or pseudo-altitudinal defects on visual field, amount of night blindness; and presence or absence of macular or optic nerve changes. These studies suggest that cone-rod degeneration patients of the RP type go through the following stages; early, the ERG has a definite cone-rod pattern where the rod ERG is larger than the cone ERG while both are abnormal. As the disease advances, there is more of a reduction in the scotopic ERG such that both the rod and cone ERGs become nearly equal. As the disease further progresses the ERG becomes non-recordable on single-flash technique, but there is good residual rod function and the final rod threshold remains good until the visual field is reduced, typically less than 10 degrees with the IV-4 isopter. Finally with advanced disease the patient becomes night blind and generally becomes very difficult to distinguished from patients who have advanced rod-cone degeneration. While it may seem logical to find that visual field size correlates with various ERG parameters; this has not been as consistent a finding in patients with rod-cone degeneration in the author's experience. The analysis shows several new pieces of information about visual field changes in cone-rod degeneration; enlarged blind spots are seen earlier in cases which have recordable cone-rod patterns (group I), and pseudo-altitudinal changes are more likely to occur in autosomal recessive patients. Patients with macular lesions and central scotomata had larger amplitudes than patients with normal appearing maculae and no central scotomata. Patients with temporal optic atrophy had an earlier onset of symptoms and significant correlation with both photopic a- and b-waves and bright flash

  7. Evaluation of neutron background in cryogenic Germanium target for WIMP direct detection when using reactor neutrino detector as neutron veto

    NASA Astrophysics Data System (ADS)

    Xu, Ye; Lan, Jieqin; Bai, Ying; Gao, Weiwei

    2016-09-01

    A direct WIMP (Weakly Interacting Massive Particle) detector with a neutron veto system is designed to better reject neutrons. An experimental configuration is studied in the present paper: 984 Ge modules are placed inside a reactor neutrino detector. In order to discriminate between nuclear and electron recoil, both ionization and heat signatures are measured using cryogenic germanium detectors in this detection. The neutrino detector is used as a neutron veto device. The neutron background for the experimental design has been estimated using the Geant4 simulation. The results show that the neutron background can decrease to O(0.01) events per year per tonne of high purity Germanium. We calculate the sensitivity to spin-independent WIMP-nucleon elastic scattering. An exposure of one tonne × year could reach a cross-section of about 2×10-11 pb.

  8. REACTIVITY MEASUREMENT FACILITY. CAMERA LOOKS DOWN INTO MTR CANAL. REACTOR ...

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

    REACTIVITY MEASUREMENT FACILITY. CAMERA LOOKS DOWN INTO MTR CANAL. REACTOR IS FUELED AS AN ETR MOCK-UP. LIGHTS DANGLE BELOW WATER LEVEL. CONTROL RODS AND OTHER APPARATUS DESCEND FROM ABOVE WATER LEVEL. INL NEGATIVE NO. 56-900. Jack L. Anderson, Photographer, 3/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  9. Neutronics Simulations of 237Np Targets to Support Safety-Basis and 238Pu Production Assessment Efforts at the High Flux Isotope Reactor

    SciTech Connect

    Chandler, David; Ellis, Ronald James

    2015-01-01

    Fueled by two highly enriched uranium-bearing fuel elements surrounded by a large concentric ring of beryllium reflector, the High Flux Isotope Reactor (HFIR) provides one of the highest neutron fluxes in the world and is used to produce unique isotopes like plutonium-238. The National Aeronautics and Space Administration use radioisotope thermoelectric generators powered by 238Pu for deep-space missions. As part of the US Department of Energy s task to reestablish the domestic production of 238Pu, a technology demonstration sub-project has been initiated to establish a new 238Pu supply chain. HFIR safety-basis neutronics calculations are being performed to ensure the target irradiations have no adverse impacts on reactor performance and to calculate data required as input to follow-on thermal-structural, thermal-hydraulic and radionuclide/dose analyses. Plutonium-238 production assessments are being performed to estimate the amount of 238Pu that can be produced in HFIR s permanent beryllium reflector. It is estimated that a total of 0.96 1.12 kg 238Pu (~1.28 1.49 kg PuO2 at 85% 238Pu/Pu purity) could be produced per year in HFIR s permanent beryllium reflector irradiation facilities if they are all utilized.

  10. 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

  11. Critical Power in 7-Rod Tight Lattice Bundle

    NASA Astrophysics Data System (ADS)

    Liu, Wei; Kureta, Masatoshi; Akimoto, Hajime

    The Reduced-Moderation Water Reactor (RMWR) has recently becomes of great concern. The RMWR is expected to promote the effective utilization of uranium recourse. The RMWR is based on water-cooled reactor technology, with achieved under lower core water volume and water flow rate. In comparison with the current light water reactors whose water-to-fuel volume ratio is about 2-3, in the RMWR, this value is reduced to less than 0.5. Thereby, there is a need to research its cooling characteristics. Experimental research on critical power in tight lattice bundle that simulates the RMWR has been carried out in Japan Atomic Energy Research Institute (JAERI). The bundle consists one center rod and six peripheral rods. The 7 rods are arranged on a 14.3mm equilateral triangular pitch. Each rod is 13mm in outside diameter. An axial 12-step power distribution is employed to simulate the complicate heating condition in RMWR. Experiments are carried out under G=100-1400kg/m2s, Pex=2-8.5MPa. Effects of mass velocity, inlet temperature, pressure, radial peaking factor and axial peaking factor on critical power and critical quality are discussed. Compared with axial uniform heating condition, the axial non-uniform heating condition causes an obvious decrease in critical quality. Arai correlation, which is the only correlation that has been optimized for tight lattice condition, is verified with the present experimental data. The correlation is found to be able to give reasonable prediction only around RMWR nominal operating condition.

  12. Transmutation of Radioactive Nuclear Waste — Present Status and Requirement for the Problem-Oriented Nuclear Database: Approach to Scheduling the Experiments (Reactor, Target, Blanket)

    NASA Astrophysics Data System (ADS)

    Artisyuk, V.; Ignatyuk, A.; Korovin, Yu.; Lopatkin, A.; Matveenko, I.; Stankovskiy, A.; Titarenko, Yu.

    2005-05-01

    Transmutation of nuclear wastes (Minor Actinides and Long-Lived Fission Products) remains an important option to reduce the burden of high-level waste on final waste disposal in deep geological structures. Accelerator-Driven Systems (ADS) are considered as possible candidates to perform transmutation due to their subcritical operation mode that eliminates some of the serious safety penalties unavoidable in critical reactors. Specific requirements to nuclear data necessary for ADS transmutation analysis is the main subject of the ISTC Project ♯2578 which started in 2004 to identify the areas of research priorities in the future. The present paper gives a summary of ongoing project stressing the importance of nuclear data for blanket performance (reactivity behavior with associated safety characteristics) and uncertainties that affect characteristics of neutron producing target.

  13. 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.

  14. Rod guide/paraffin scraper

    SciTech Connect

    Mabry, J.F.

    1991-02-26

    This patent describes improvement in a rod guide and paraffin scraper. It comprises: a body including longitudinal ribs spaced radially and extending out from the body; having two identical halves with the body surrounding a bore to accept a sucker rod, and each of the identical halves having a locking and tightening feature using a tongue and groove concept for interfitting the halves together over the sucker rod. This improvement comprises a rod guide and paraffin scraper with two identical halves comprising; a cylindrical central body including, at each end, three longitudinal ribs radially spaced to form a triad leaving three flow channels, at each end of the body, of essentially the same size and spacing as the ribs; and an angular wedge with opposingly ramped sides at the inside end of each of the ribs for scraping and directing material into the flow channels; and a set of triangular shaped tongues that interfit with a set of triangular shaped grooves for tightening the identical halves together and over the sucker rod; and a pair of cone-shaped male locks at one end of the identical half to mate with a pair of cone-shaped female locks at the opposite end of the other identical half.

  15. A survey of control rod measurements in ZPPR and their analysis

    SciTech Connect

    Collins, P.J.

    1988-01-01

    The accurate prediction of control rod worths has been of great concern in the United States. Optimum control configurations need to balance several often conflicting requirements of control through the operating cycle, while maintaining acceptable power shapes, safety considerations of overriding importance, together with seeking economy by minimizing the number of rods, reducing boron enrichment and lengthening replacement intervals. After control and shutdown requirements have been met, the most important safety concern is the transient overpower condition (TOP) which may be initiated by uncontrolled run-out of a primary rod. Stringent criteria for the primary and secondary systems may be that they are independently capable of shutting down the reactor even with one rod stuck. The TOP initiator may be greatly enhanced by control rod interaction effects. Control rod effects may have a strong impact on core design. For example, work on the integral fast reactor with metallic fuel at ANL has studied core designs which minimize the TOP reactivity by maintaining a minimum primary control bank insertion through tailoring the internal breeding gain. The predicted control rod worths are very sensitive to the calculation methods used and to the accuracy of the basic nuclear data files. Required accuracies have been achieved only through the use of critical experiments on the ZPR and ZPPR facilities. Experiments on ZPR-3 and ZPR-9 produced satisfactory control predictions for the SEFOR, EBR-II and FFTF reactors. This document provides a survey of control rod measurements and compares calculated and experimental results. 16 refs., 3 figs., 10 tabs.

  16. 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).

  17. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  18. 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

  19. SYSTEM FOR UNLOADING REACTORS

    DOEpatents

    Rand, A.C. Jr.

    1961-05-01

    An unloading device for individual vertical fuel channels in a nuclear reactor is shown. The channels are arranged in parallel rows and underneath each is a separate supporting block on which the fuel in the channel rests. The blocks are raounted in contiguous rows on an array of parallel pairs of tracks over the bottom of the reactor. Oblong hollows in the blocks form a continuous passageway through the middle of the row of blocks on each pair of tracks. At the end of each passageway is a horizontal grappling rod with a T- or L extension at the end next to the reactor of a length to permit it to pass through the oblong passageway in one position, but when rotated ninety degrees the head will strike one of the longer sides of the oblong hollow of one of the blocks. The grappling rod is actuated by a controllable reciprocating and rotating device which extends it beyond any individual block desired, rotates it and retracts it far enough to permit the fuel in the vertical channel above the block to fall into a handling tank below the reactor.

  20. Application of BACE1 immobilized enzyme reactor for the characterization of multifunctional alkaloids from Corydalis cava (Fumariaceae) as Alzheimer's disease targets.

    PubMed

    Chlebek, Jakub; De Simone, Angela; Hošťálková, Anna; Opletal, Lubomír; Pérez, Concepción; Pérez, Daniel I; Havlíková, Lucie; Cahlíková, Lucie; Andrisano, Vincenza

    2016-03-01

    In our ongoing study focused on Corydalis cava (Fumariaceae), used in folk medicine in the treatment of memory dysfunctions, we have investigated fifteen previously isolated alkaloids for their potential multifunctional activity on Alzheimer's disease (AD) targets. Determination of ß-site amyloid precursor protein cleaving enzyme 1 (BACE1) inhibition was carried out using a BACE1-Immobilized Enzyme Reactor (IMER) by validating the assay with a multi-well plate format Fluorescence Resonance Energy Transfer (FRET) assay. Seven alkaloids out of fifteen were found to be active, with (-)-corycavamine (3) and (+)-corynoline (5) demonstrating the highest BACE1 inhibition activity, in the micromolar range, in a concentration dependent manner. BACE1-IMER was found to be a valid device for the fast screening of inhibitors and the determination of their potency. In a permeation assay (PAMPA) for the prediction of blood-brain barrier (BBB) penetration, the most active compounds, (-)-corycavamine (3) and (+)-corynoline (5), were found to be able to cross the BBB. Not all compounds showed activity against glycogen synthase kinase-3β (GSK-3β) and casein kinase-1δ (CK-1δ). On the basis of the reported results, we found that some C. cava alkaloids have multifunctional activity against AD targets (prolyl oligopeptidase, cholinesterases and BACE1). Moreover, we tried to elucidate the treatment effectivity (rational use) of its extract in memory dysfunction in folk medicine. PMID:26779945

  1. Application of BACE1 immobilized enzyme reactor for the characterization of multifunctional alkaloids from Corydalis cava (Fumariaceae) as Alzheimer's disease targets.

    PubMed

    Chlebek, Jakub; De Simone, Angela; Hošťálková, Anna; Opletal, Lubomír; Pérez, Concepción; Pérez, Daniel I; Havlíková, Lucie; Cahlíková, Lucie; Andrisano, Vincenza

    2016-03-01

    In our ongoing study focused on Corydalis cava (Fumariaceae), used in folk medicine in the treatment of memory dysfunctions, we have investigated fifteen previously isolated alkaloids for their potential multifunctional activity on Alzheimer's disease (AD) targets. Determination of ß-site amyloid precursor protein cleaving enzyme 1 (BACE1) inhibition was carried out using a BACE1-Immobilized Enzyme Reactor (IMER) by validating the assay with a multi-well plate format Fluorescence Resonance Energy Transfer (FRET) assay. Seven alkaloids out of fifteen were found to be active, with (-)-corycavamine (3) and (+)-corynoline (5) demonstrating the highest BACE1 inhibition activity, in the micromolar range, in a concentration dependent manner. BACE1-IMER was found to be a valid device for the fast screening of inhibitors and the determination of their potency. In a permeation assay (PAMPA) for the prediction of blood-brain barrier (BBB) penetration, the most active compounds, (-)-corycavamine (3) and (+)-corynoline (5), were found to be able to cross the BBB. Not all compounds showed activity against glycogen synthase kinase-3β (GSK-3β) and casein kinase-1δ (CK-1δ). On the basis of the reported results, we found that some C. cava alkaloids have multifunctional activity against AD targets (prolyl oligopeptidase, cholinesterases and BACE1). Moreover, we tried to elucidate the treatment effectivity (rational use) of its extract in memory dysfunction in folk medicine.

  2. Analysis of reciprocating compressor piston rod failures

    SciTech Connect

    Tripp, H.A.; Drosjack, M.J.

    1984-02-01

    This report presents the analysis of five piston rod failures which occurred on reciprocating compressors. Calculations are shown for rod stress which includes nominal rod loading sources as well as additional loads due to unusual pressure losses in the compressor valves, flexure of the rods due to misalignment, and manufacturing errors. The additional loads were incorporated on the basis of field measurements. The stress values are used with Baquin's equation to produce fatigue life curves for the rods. Based on the calculations, recommendations for modified rods were made. The calculation procedures are described in a manner which will permit their application to other reciprocating compressors.

  3. Materials Test-2 LOCA Simulation in the NRU Reactor

    SciTech Connect

    Barner, J. O.; Hesson, G. M.; King, I. L.; Marshall, R. K.; Parchen, L. J.; Pilger, J. P.; Rausch, W. N.; Russcher, G. E.; Webb, B. J.; Wildung, N. J.; Wilson, C. L.; Wismer, M. D.; Mohr, C. L.

    1982-03-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This third experiment of the program produced fuel cladding temperatures exceeding 1033 K (1400°F) for 155 s and resulted in eight ruptured fuel rods. Experiment data and initial results are presented in the form of photographs and graphical summaries.

  4. Demonstration of EBR-II power maneuvers without control rod movement

    SciTech Connect

    Chang, L.K.; Mohr, D.; Planchon, H.P.; Feldman, E.E.; Messick, N.C.

    1988-01-01

    A group of five plant inherent control tests was successfully conducted in November 1987 in the Experimental Breeder Reactor II. These tests demonstrated that the plant power of a metal-fueled reactor can be passively controlled over a large power range by slowly changing the primary flow and the reactor inlet temperature. These variables are, in turn, regulated by the primary pump speed, the secondary flow, and the turbine inlet pressure. In all tests, control rods were not used to regulate power. It was demonstrated that the plant power can be controlled with reasonable accuracy without using control rods when the reactivity feedback characteristics of the reactor are well understood and the plant controllers are adequately designed.

  5. Application of fiberglass sucker rods

    SciTech Connect

    Gibbs, S.G. )

    1991-05-01

    Fiberglass sucker rods are assuming a place in artificial-lift technology. This paper briefly describes the manufacturing process and gives some design and operational hints for practical applications. It also describes some mathematical modeling modifications needed for fiberglass wave-equation design programs.

  6. 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.

  7. What operators say about fiberglass sucker rods

    SciTech Connect

    Bleakley, W.B.

    1984-11-01

    This article presents the results of an informal survey of oil producing companies and one design engineering firm in the Permian Basin about the use and performance of fiberglass sucker rods in sucker rod pumps.

  8. 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).

  9. 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.

  10. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  11. Moon base reactor system

    NASA Technical Reports Server (NTRS)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  12. RADIATION FACILITY FOR NUCLEAR REACTORS

    DOEpatents

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

    1961-12-12

    A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

  13. BWR feedwater nozzle and control-rod-drive return line nozzle cracking

    SciTech Connect

    Not Available

    1981-11-01

    In its 1978 Annual Report to Congress, the Nuclear Regulatory Commission identified as an unresolved safety issue the appearance of cracks in feedwater nozzles at boiling-water reactors (BWRs). Later similar cracking, detected in return water lines for control-rod-drive systems at BWRs, was designated Part II of the issue. This article outlines the resolution of these cracking problems.

  14. Parallel Magnetic Flow Electromagnet for Movable Coil Control-rod Driving Mechanism

    SciTech Connect

    Jige, Zhang

    2006-07-01

    The parallel magnetic flow electromagnet can effectively relax the saturation, which easily takes place in the single magnetic flow electromagnet, and accordingly can improve the drive capacity of the movable coil electromagnet drive mechanism for a mobile reactor control rod. (authors)

  15. Who makes API sucker rods and couplings

    SciTech Connect

    Not Available

    1986-03-01

    This guide identifies manufacturers qualified to produce API sucker rods and related equipment, lists chemical and mechanical properties of the various types of rods and provides dimensional characteristics. In addition, similar information is given for non-API rods such as fiberglass and aluminum.

  16. 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 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. 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 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...

  18. 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 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...

  19. 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 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...

  20. 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...

  1. Passive Reactor Cooling Using Capillary Porous Wick

    SciTech Connect

    Miller, Christopher G.; Lin, Thomas F.

    2006-07-01

    Long-term reliability of actively pumped cooling systems is a concern in space-based nuclear reactors. Capillary-driven passive cooling systems are being considered as an alternative to gravity-driven systems. The high surface tension of liquid lithium makes it attractive as the coolant in a capillary-driven cooling system. A system has been conceived in which the fuel rod of a reactor is surrounded by a concentric wick through which liquid lithium flows to provide cooling under normal and emergency operating conditions. Unheated wicking experiments at three pressures using four layered screen mesh wicks of different porosities and three relatively high surface tension fluids have been conducted to gain insight into capillary phenomena for such a capillary cooling system. All fluids tested demonstrated wicking ability in each of the wick structures for all pressures, and wicking ability for each fluid increased with decreasing wick pore size. An externally heated wicking experiment with liquid lithium as the wicking fluid was also conducted. In addition to wicking experiments, a heater rod is under development to simulate the fuel rod of a space based nuclear reactor by providing a heat flux of up to 110 kW/m{sup 2}. Testing of this heater rod has shown its ability to undergo repeated cycling from below 533 K to over 1255 K without failure. This heater rod will be integrated into lithium wicking experiments to provide more realistic simulation of the proposed capillary-driven space nuclear reactor cooling system. (authors)

  2. Mechanics of high speed impact at normal incidence between plasticine long rods and plates

    NASA Astrophysics Data System (ADS)

    Johnson, W.; Sengupta, A. K.; Ghosh, S. K.; Reid, S. R.

    1981-12-01

    A BRIEF literature review of long rod impact is given. Experimental results and observations on the penetration and perforation of plasticine targets of finite thickness (25 and 50 mm) impinged upon normally by long rods of plasticine of various length-to-diameter ratios (2-12) are presented and discussed. An account of the range of modes of deformation as it pertains to long rods and targets of identical materials is given, i.e. mushrooming of the projectile leading-end, partial penetration and embedment of the projectile in the target, gross target penetration by the projectile, and penetration followed by perforation. The difference in penetration behaviour as between long and short rods is discussed. Plasticine has been used for both the target and projectile materials in order to simulate hypervelocity impact conditions for deformable long rod projectiles on deformable target plates. This permits the use of equipment based on an industrial stud driver which is substantially cheaper than the equipment (e.g. a light gas gun) required to achieve the range of speeds necessary for hypervelocity conditions with real ballistics-related metals and alloys. Projectile-target interactions and damage observed in the present tests compare well with those reported in similar investigations using metallic projectiles and targets. The attraction of performing ballistics experiments with plasticine models outweighs such inherent limitations as the inability to model the extreme conditions of temperature and strain-hardening in real materials, because testing is cheap and easily performed in a laboratory.

  3. 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.

  4. 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.

  5. 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.

  6. Hanging core support system for a nuclear reactor. [LMFBR

    DOEpatents

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

    1984-04-26

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

  7. Guide for rotating sucker rods

    SciTech Connect

    Harrel, R.D.

    1986-11-04

    This patent describes an improved guide for use in a string of sucker rods rotated in a tubing string in a borehole, the sucker rods having threaded male ends, the guide comprising: an elongated upright solid cylindrical coupling body of external diameter less than the internal diameter of tubing in which it is to be used; a pair of spaced apart axle holders positioned in three recess; an axle received in each recess in the coupling body, the axis of each axle being parallel and spaced from the body longitudinal axis; a roller rotatably received on each axle, the periphery of each roller extending exteriorly of the external cylindrical surface of the coupling body; and means to retain each of the holders in the coupling body recess.

  8. 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).

  9. The rod circuit in the rabbit retina.

    PubMed

    Vaney, D I; Young, H M; Gynther, I C

    1991-01-01

    Mammalian retinae have a well-defined neuronal pathway that serves rod vision. In rabbit retina, the different populations of interneurons in the rod pathway can be selectively labeled, either separately or in combination. The rod bipolar cells show protein kinase C immunoreactivity; the rod (AII) amacrine cells can be distinguished in nuclear-yellow labeled retina; the rod reciprocal (S1 & S2) amacrine cells accumulate serotonin; and the dopaminergic amacrine cells show tyrosine-hydroxylase immunoreactivity. Furthermore, intracellular dye injection of the microscopically identified interneurons enables whole-population and single-cell studies to be combined in the same tissue. Using this approach, we have been able to analyze systematically the neuronal architecture of the rod circuit across the rabbit retina and compare its organization with that of the rod circuit in central cat retina. In rabbit retina, the rod interneurons are not organized in a uniform neuronal module that is simply scaled up from central to peripheral retina. Moreover, peripheral fields in superior and inferior retina that have equivalent densities of each neuronal type show markedly different rod bipolar to AII amacrine convergence ratios, with the result that many more rod photoreceptors converge on an AII amacrine cell in superior retina. In rabbit retina, much of the convergence in the rod circuit occurs in the outer retina whereas, in central cat retina, it is more evenly distributed between the inner and outer retina.

  10. Tests pinpoint sucker-rod failures

    SciTech Connect

    Elshawesh, F.; Elhoud, A.; Elagdel, E.

    1997-05-26

    A detailed metallurgical examination of a 7/8-inch and a 1-inch sucker rod revealed corrosion fatigue had caused their failure. The 7 to 8-inch rod had failed after a few months of service while the 1-inch rod failed after 1 year. Both rods had been used in a sweet-oil environment. Both rods failed by corrosion fatigue because of repeated loads during operations. Pitting because of the presence of chloride ions and carbon dioxide was initiated on the rod surface, which in turn acted as a crack origin from which the fatigue crack initiated and propagated during operations. The pitting was on the external surface. These pits were large and penetrated through the rod cross-section. Fatigue cracking is initiated at the bottom of the pit where high stress concentration is expected and propagated because the rods were subjected to the alternating stresses during operation. The extent of the fatigue crack varied in the two examined rods because of the difference in the rod heat treatment and microstructure. The paper discusses fatigue failure, the visual examination, macroscopic and microscopic examinations, rod properties, and future operations.

  11. 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.

  12. Root cause of incomplete control rod insertions at Westinghouse reactors

    SciTech Connect

    Ray, S.

    1997-01-01

    Within the past year, incomplete RCCA insertions have been observed on high burnup fuel assemblies at two Westinghouse PWRs. Initial tests at the Wolf Creek site indicated that the direct cause of the incomplete insertions observed at Wolf Creek was excessive fuel assembly thimble tube distortion. Westinghouse committed to the NRC to perform a root cause analysis by the end of August, 1996. The root cause analysis process used by Westinghouse included testing at ten sites to obtain drag, growth and other characteristics of high burnup fuel assemblies. It also included testing at the Westinghouse hot cell of two of the Wolf Creek incomplete insertion assemblies. A mechanical model was developed to calculate the response of fuel assemblies when subjected to compressive loads. Detailed manufacturing reviews were conducted to determine if this was a manufacturing related issue. In addition, a review of available worldwide experience was performed. Based on the above, it was concluded that the thimble tube distortion observed on the Wolf Creek incomplete insertion assemblies was caused by unusual fuel assembly growth over and above what would typically be expected as a result of irradiation exposure. It was determined that the unusual growth component is a combination of growth due to oxide accumulation and accelerated growth, and would only be expected in high temperature plants on fuel assemblies that see long residence times and high power duties.

  13. Calcium spikes in toad rods.

    PubMed Central

    Fain, G L; Gerschenfeld, H M; Quandt, F N

    1980-01-01

    1. When the retina of the toad, Bufo marinus, was superfused with 6-12 mM-tetraethylammonium chloride (TEA), intracellular recordings from rods showed large, depolarizing regenerative potentials. For brief exposures to TEA, these potentials occurred during the recovery phase of the light responses; whereas, during longer exposures, they were spontaneous in darkness but suppressed during illumination. Similar regenerative potentials were observed during perfusion with 3-10 mM-4-aminopyridine and 1-2 mM-BaCl2. 2. The amplitude of the regenerative potentials depended upon the extracellular Ca concentration ([Ca2+]o). Lowering [Ca2+]o decreased their amplitude and in zero [Ca2+]o they were reversibly abolished. Increasing [Ca2+]o by 1.5-2 times produced a small hyperpolarization of membrane potential and a large augmentation in regenerative response amplitude. However, larger increases in [Ca2+]o produced large membrane hyperpolarizations and reversibly suppressed the regenerative responses. 3. High concentrations of Sr2+ in TEA also enhanced regenerative activity but did not affect the rod resting membrane potential. The amplitude of regenerative potentials increased continuously with increasing [Sr2+]o, and in 28 mM-Sr2+ the rods generated 60-70 mV action potentials, even in the absence of extracellular Na+. 4. The regenerative potentials were blocked by 25 microM-Cd2+, 50-100 microM-Co2+, 5mM-Mg2+, and 100 microM-D-600. They were unaffected by 2 microM-TTX or 2-5 mM-Na aspartate. 5. In Ringer containing 12 mM-TEA, large anode break responses could be recorded from rods at the termination of inward current pulses. These anode break responses were also suppressed by Co2+ and unaffected by TTX or Na aspartate. 6. We conclude that the membrane of toad rods contains a conductance normally selective for Ca2+, which is activated by depolarization. In normal Ringer, the inward current through this conductance produces little effect, since it is balanced by a large outward

  14. NEUTRONIC REACTOR CONSTRUCTION

    DOEpatents

    Vernon, H.C.; Goett, J.J.

    1958-09-01

    A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

  15. Reactor refueling containment system

    DOEpatents

    Gillett, James E.; Meuschke, Robert E.

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  16. Reactor refueling containment system

    DOEpatents

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  17. Gene expression changes during retinal development and rod specification

    PubMed Central

    Carrigan, Matthew; Hokamp, Karsten; Farrar, G. Jane

    2015-01-01

    Purpose Retinitis pigmentosa (RP) typically results from individual mutations in any one of >70 genes that cause rod photoreceptor cells to degenerate prematurely, eventually resulting in blindness. Gene therapies targeting individual RP genes have shown efficacy at clinical trial; however, these therapies require the surviving photoreceptor cells to be viable and functional, and may be economically feasible for only the more commonly mutated genes. An alternative potential treatment strategy, particularly for late stage disease, may involve stem cell transplants into the photoreceptor layer of the retina. Rod progenitors from postnatal mouse retinas can be transplanted and can form photoreceptors in recipient adult retinas; optimal numbers of transplantable cells are obtained from postnatal day 3–5 (P3–5) retinas. These cells can also be expanded in culture; however, this results in the loss of photoreceptor potential. Gene expression differences between postnatal retinas, cultured retinal progenitor cells (RPCs), and rod photoreceptor precursors were investigated to identify gene expression patterns involved in the specification of rod photoreceptors. Methods Microarrays were used to investigate differences in gene expression between cultured RPCs that have lost photoreceptor potential, P1 retinas, and fresh P5 retinas that contain significant numbers of transplantable photoreceptors. Additionally, fluorescence-activated cell sorting (FACS) sorted Rho-eGFP-expressing rod photoreceptor precursors were compared with Rho-eGFP-negative cells from the same P5 retinas. Differential expression was confirmed with quantitative polymerase chain reaction (q-PCR). Results Analysis of the microarray data sets, including the use of t-distributed stochastic neighbor embedding (t-SNE) to identify expression pattern neighbors of key photoreceptor specific genes, resulted in the identification of 636 genes differentially regulated during rod specification. Forty-four of these

  18. Model of ASTM Flammability Test in Microgravity: Iron Rods

    NASA Technical Reports Server (NTRS)

    Steinberg, Theodore A; Stoltzfus, Joel M.; Fries, Joseph (Technical Monitor)

    2000-01-01

    There is extensive qualitative results from burning metallic materials in a NASA/ASTM flammability test system in normal gravity. However, this data was shown to be inconclusive for applications involving oxygen-enriched atmospheres under microgravity conditions by conducting tests using the 2.2-second Lewis Research Center (LeRC) Drop Tower. Data from neither type of test has been reduced to fundamental kinetic and dynamic systems parameters. This paper reports the initial model analysis for burning iron rods under microgravity conditions using data obtained at the LERC tower and modeling the burning system after ignition. Under the conditions of the test the burning mass regresses up the rod to be detached upon deceleration at the end of the drop. The model describes the burning system as a semi-batch, well-mixed reactor with product accumulation only. This model is consistent with the 2.0-second duration of the test. Transient temperature and pressure measurements are made on the chamber volume. The rod solid-liquid interface melting rate is obtained from film records. The model consists of a set of 17 non-linear, first-order differential equations which are solved using MATLAB. This analysis confirms that a first-order rate, in oxygen concentration, is consistent for the iron-oxygen kinetic reaction. An apparent activation energy of 246.8 kJ/mol is consistent for this model.

  19. CRC handbook of nuclear reactors calculations. Vol. III

    SciTech Connect

    Ronen, Y.

    1986-01-01

    This handbook breaks down the complex field of nuclear reactor calculations into major steps. Each step presents a detailed analysis of the problems to be solved, the parameters involved, and the elaborate computer programs developed to perform the calculations. This book bridges the gap between nuclear reactor theory and the implementation of that theory, including the problems to be encountered and the level of confidence that should be given to the methods described. Volume III: Control Rods and Burnable Absorber Calculations. Perturbation Theory for Nuclear Reactor Analysis. Thermal Reactors Calculations. Fast Reactor Calculations. Seed-Blanket Reactors. Index.

  20. Self-sensing CF-GFRP rods as mechanical reinforcement and sensors of concrete beams

    NASA Astrophysics Data System (ADS)

    Nanni, F.; Auricchio, F.; Sarchi, F.; Forte, G.; Gusmano, G.

    2006-02-01

    In this paper testing carried out on concrete beams reinforced with self-sensing composite rods is presented. Such concrete beams, whose peculiarity is to be reinforced by self-sensing materials able to generate an alarm signal when fixed loads are reached, were designed, manufactured and tested. The reinforcing rods were manufactured by pultrusion and consisted of self-sensing hybrid composites containing both glass and carbon fibres in an epoxy resin. The experimentation was carried out by performing simultaneously mechanical tests on the reinforced beams and electrical measurements on the composite rods. The results showed that the developed system reached the target proposed, giving an alarm signal.

  1. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  2. Mechanism for selective synaptic wiring of rod photoreceptors into the retinal circuitry and its role in vision

    PubMed Central

    Cao, Yan; Sarria, Ignacio; Fehlhaber, Katherine E.; Kamasawa, Naomi; Orlandi, Cesare; James, Kiely N.; Hazen, Jennifer L.; Gardner, Matthew R.; Farzan, Michael; Lee, Amy; Baker, Sheila; Baldwin, Kristin; Sampath, Alapakkam P.; Martemyanov, Kirill A.

    2015-01-01

    SUMMARY In the retina, rod and cone photoreceptors form distinct connections with different classes of downstream bipolar cells. However, the molecular mechanisms responsible for their selective connectivity are unknown. Here we identify a cell-adhesion protein, ELFN1, to be essential for the formation of synapses between rods and rod ON-bipolar cells in the primary rod pathway. ELFN1 is expressed selectively in rods where it is targeted to the axonal terminals by the synaptic release machinery. At the synapse, ELFN1 binds in trans to mGluR6, the postsynaptic receptor on rod ON-bipolar cells. Elimination of ELFN1 in mice prevents the formation of synaptic contacts involving rods, but not cones, allowing a dissection of the contributions of primary and secondary rod pathways to retinal circuit function and vision. We conclude that ELFN1 is necessary for the selective wiring of rods into the primary rod pathway and is required for high sensitivity of vision. PMID:26402607

  3. ENGINEERING TEST REACTOR

    DOEpatents

    De Boisblanc, D.R.; Thomas, M.E.; Jones, R.M.; Hanson, G.H.

    1958-10-21

    Heterogeneous reactors of the type which is both cooled and moderated by the same fluid, preferably water, and employs highly enriched fuel are reported. In this design, an inner pressure vessel is located within a main outer pressure vessel. The reactor core and its surrounding reflector are disposed in the inner pressure vessel which in turn is surrounded by a thermal shield, Coolant fluid enters the main pressure vessel, fiows downward into the inner vessel where it passes through the core containing tbe fissionable fuel assemblies and control rods, through the reflector, thence out through the bottom of the inner vessel and up past the thermal shield to the discharge port in the main vessel. The fuel assemblles are arranged in the core in the form of a cross having an opening extending therethrough to serve as a high fast flux test facility.

  4. AIR COOLED NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

  5. Spring element for holding down nuclear reactor fuel assembly

    SciTech Connect

    Steinke, A.

    1981-07-14

    Spring element is described for holding down and bracing a fuel assembly against a hold-down plate upwardly limiting the reactor core of a nuclear reactor. Includes a spring-loaded rod-shaped member separately formed independently of the fuel assembly and being slidable axially and form-lockingly into the fuel assembly.

  6. Analysis of the Reactor Position Independent Monitor (PIM) Diagnostic

    SciTech Connect

    Hayes-Sterbenz, Anna Catherine

    2014-07-17

    In this note I analyze the physics determining the proposed reactor position independent monitor (PIM), which is the ratio (240Pu/239Pu)1/3 × (135Cs/137Cs)1/2. The PIM ratios in any reactor fuel is shown to increase monotonically with the time over which the fuel is irradiated. This is because the Cs ratio determines the neutron flux, while the Pu isotopic ratio is determined by the flux times the irradiation time. If the irradiation time for all fuel rods across the reactor is fixed, the PIM ratio is approximately constant in all rods. However, no information can be extracted from the PIM ratio on Pu isotopics unless both the flux (or Cs ratio) and the irradiation time (from, say, Ru isotopics) are known separately, i.e., the PIM ratio is not a fundamental parameter of any reactor. Thus, unless the PIM ratio has been measured for the specific fuel under interrogation, no information can be deduced from measurements or reactor simulations of PIM ratios in different fuel from the same reactor. However, if a PIM measurement has been in one spent fuel rod from a given reactor, all other rods that are known to have been in the reactor for the same irradiation period can be assumed to have approximately the same PIM ratio.

  7. Materials considerations in accelerator targets

    SciTech Connect

    Peacock, H.B. Jr.; Iyer, N.C.; Louthan, M.R. Jr.

    1994-08-01

    Future nuclear materials production and/or the burn-up of long lived radioisotopes may be accomplished through the capture of spallation produced neutrons in accelerators. Aluminum clad-lead and/or lead alloys has been proposed as a spallation target. Aluminum was the cladding choice because of the low neutron absorption cross section, fast radioactivity decay, high thermal conductivity, and excellent fabricability. Metallic lead and lead oxide powders were considered for the target core with the fabrication options being casting or powder metallurgy (PM). Scoping tests to evaluate gravity casting, squeeze casting, and casting and swaging processes showed that, based on fabricability and heat transfer considerations, squeeze casting was the preferred option for manufacture of targets with initial core cladding contact. Thousands of aluminum clad aluminum-lithium alloy core targets and control rods for tritium production have been fabricated by coextrusion processes and successfully irradiated in the SRS reactors. Tritium retention in, and release from the coextruded product was modeled from experimental and operational data. Newly produced tritium atoms were trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability was the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release was determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. The model can be used to calculate tritium release from aluminum clad, aluminum-lithium alloy targets during postulated accelerator operational and accident conditions. This paper describes the manufacturing technologies evaluated and presents the model for tritium retention in aluminum clad, aluminum-lithium alloy tritium production targets.

  8. UO 2/Zry-4 chemical interaction layers for intact and leak PWR fuel rods

    NASA Astrophysics Data System (ADS)

    Kim, Kyu-Tae

    2010-09-01

    In this study, the UO 2 pellet-Zry-4 cladding interfaces of intact and leak PWR fuel rods were examined with the help of an optical microscope and a scanning electron microscope to investigate typical chemical interaction layers formed at the pellet-cladding interface during the normal reactor operations. The two intact and two leak fuel rods with the burnup of between 35,000 and 53,000 MWD/MTU were selected to evaluate the effects of gap-gas compositions and fuel burnup on the chemical interaction layer formation. Based on the optical and scanning electron micrographs, it is found that the intact fuel rod generates apparently one interaction layer of (U,Zr)O 2-x at the interface, whereas the leak fuel rod generates apparently two interaction layers of ZrO 2-x and (U,Zr)O 2-x. These interaction layers for the intact and leak fuel rods were predicted by several diffusion paths drawn on a U-Zr-O ternary phase diagram. The variations of chemical element compositions around the interface of one intact rod were generated by an electron probe micro-analyzer to confirm the interaction layers at the pellet-cladding interface. The interaction layer growth rates of the ZrO 2-x and (U,Zr)O 2-x phases were estimated, using the layer thicknesses and the reaction times.

  9. Spontaneous Patterning of Confined Granular Rods

    NASA Astrophysics Data System (ADS)

    Galanis, Jennifer; Harries, Daniel; Sackett, Dan L.; Losert, Wolfgang; Nossal, Ralph

    2006-01-01

    Vertically vibrated rod-shaped granular materials confined to quasi-2D containers self-organize into distinct patterns. We find, consistent with theory and simulation, a density dependent isotropic-nematic transition. Along the walls, rods interact sterically to form a wetting layer. For high rod densities, complex patterns emerge as a result of competition between bulk and boundary alignment. A continuum elastic energy accounting for nematic distortion and local wall anchoring reproduces the structures seen experimentally.

  10. Nuclear reactor with low-level core coolant intake

    DOEpatents

    Challberg, Roy C.; Townsend, Harold E.

    1993-01-01

    A natural-circulation boiling-water reactor has skirts extending downward from control rod guide tubes to about 10 centimeters from the reactor vessel bottom. The skirts define annular channels about control rod drive housings that extend through the reactor vessel bottom. Recirculating water is forced in through the low-level entrances to these channels, sweeping bottom water into the channels in the process. The sweeping action prevents cooler water from accumulating at the bottom. This in turn minimizes thermal shock to bottom-dwelling components as would occur when accumulated cool water is swept away and suddenly replaced by warmer water.

  11. 27. The top of a typical pile, F Reactor in ...

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

    27. The top of a typical pile, F Reactor in February 1945 in this case, showing the vertical safety rods (VSRs) and the cables that support them. The rods could be dropped into the pile to effect a rapid shutdown. The four silvered-colored drums on the left contained boron solution and are part of the last ditch safety system. Should the VSRs channels become blocked by an occurrence such as an earthquake, the solution could be dumped into the VSR channels to help shut down the reactor. D-8334 - B Reactor, Richland, Benton County, WA

  12. 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

  13. Improved model for sucker rod pumping

    SciTech Connect

    Doty, D.R.; Schmidt, Z.

    1981-01-01

    An improved model for predicting the behavior of sucker rod pumping installations is presented. This model incorporates the dynamics of the liquid columns as well as the sucker rod string through a system of partial differential equations. The system of equations is solved by a modified method of characteristics on a digital computer. The model predicts the polished rod and pump dynamometer cards and incorporates the effects of liquid inertia and viscosity. It is capable of simulating a wide variety of pumping conditions where liquid physical properties are important. The information predicted by the model is useful in the design and operation of sucker rod pumping installations. Refs.

  14. An improved model for sucker rod pumping

    SciTech Connect

    Doty, D.R.; Schmidt, Z.

    1983-02-01

    An improved model for predicting the behavior of sucker rod pumping installations is presented. This model incorporates the dynamics of the liquid columns as well as the sucker rod string through a system of partial differential equations. This system of equations is solved by a modified method of characteristics on a digital computer. The model predicts the polished-rod and pump dynamometer cards and incorporates the effects of liquid inertia and viscosity. The model is capable of simulating a wide variety of pumping conditions for which liquid physical properties are important. The information predicted by the model is useful in the design and operation of sucker rod pumping installations.

  15. 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.

  16. 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

  17. MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

    DOEpatents

    Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

    1962-06-26

    A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

  18. Cyclic Nucleotide-Gated Channels Require Ankyrin-G for Transport to the Sensory Cilium of Rod Photoreceptors

    PubMed Central

    Kizhatil, Krishnakumar; Baker, Sheila A.; Arshavsky, Vadim Y.; Bennett, Vann

    2009-01-01

    Cyclic nucleotide-gated channels localize exclusively to the plasma membrane of photosensitive outer segments of rod photoreceptors where they generate the electrical response to light. Here we found that targeting of cyclic nucleotide-gated channels to the rod outer segment required their interaction with ankyrin-G. Ankyrin-G localized exclusively to rod outer segments, coimmunoprecipitated with the cyclic nucleotide-gated channel, and bound to the C-terminal domain of the β1-subunit. Ankyrin-G depletion in neonatal mouse retinas markedly reduced cyclic nucleotide-gated channel expression. Transgenic expression of cyclic nucleotide-gated channel β-subunit mutants in Xenopus rods showed that ankyrin-G binding was necessary and sufficient for targeting of the β1-subunit to outer segments. Thus ankyrin-G is required for transport of cyclic nucleotide-gated channels to the plasma membrane of rod outer segments. PMID:19299621

  19. Benchmark Evaluation of the HTR-PROTEUS Absorber Rod Worths (Core 4)

    SciTech Connect

    John D. Bess; Leland M. Montierth

    2014-06-01

    PROTEUS was a zero-power research reactor at the Paul Scherrer Institute (PSI) in Switzerland. The critical assembly was constructed from a large graphite annulus surrounding a central cylindrical cavity. Various experimental programs were investigated in PROTEUS; during the years 1992 through 1996, it was configured as a pebble-bed reactor and designated HTR-PROTEUS. Various critical configurations were assembled with each accompanied by an assortment of reactor physics experiments including differential and integral absorber rod measurements, kinetics, reaction rate distributions, water ingress effects, and small sample reactivity effects [1]. Four benchmark reports were previously prepared and included in the March 2013 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [2] evaluating eleven critical configurations. A summary of that effort was previously provided [3] and an analysis of absorber rod worth measurements for Cores 9 and 10 have been performed prior to this analysis and included in PROTEUS-GCR-EXP-004 [4]. In the current benchmark effort, absorber rod worths measured for Core Configuration 4, which was the only core with a randomly-packed pebble loading, have been evaluated for inclusion as a revision to the HTR-PROTEUS benchmark report PROTEUS-GCR-EXP-002.

  20. 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

  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.

  2. 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.

  3. Sucker rod makers offer a selection

    SciTech Connect

    Savage, D.

    1983-11-01

    In their ongoing effort to produce better, more cost-effective sucker rods, manufacturers have selected one of three materials - fiberglass, aluminum, and steel - that they feel best suits the production system function of the rods, which is to connect the downhole pump to the pumpjack on the surface. Characteristics of each are described.

  4. Longitudinal Impact of Rods: A Continuing Experiment.

    ERIC Educational Resources Information Center

    Britton, W. G. B.; And Others

    1978-01-01

    Describes an undergraduate experiment of research potential. The experiment cconsists of measuring the time of contact of a metal rod bouncing on a steel base as a function of the velocity of impact, length, diameter, and material of the rod. (GA)

  5. 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…

  6. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  7. Degradation in steam of 60 cm-long B4C control rods

    NASA Astrophysics Data System (ADS)

    Dominguez, C.; Drouan, D.

    2014-08-01

    In the framework of nuclear reactor core meltdown accident studies, the degradation of boron carbide control rod segments exposed to argon/steam atmospheres was investigated up to about 2000 °C in IRSN laboratories. The sequence of the phenomena involved in the degradation has been found to take place as expected. Nevertheless, the ZrO2 oxide layer formed on the outer surface of the guide tube was very protective, significantly delaying and limiting the guide tube failure and therefore the boron carbide pellet oxidation. Contrary to what was expected, the presence of the control rod decreases the hydrogen release instead of increasing it by additional oxidation of boron compounds. Boron contents up to 20 wt.% were measured in metallic mixtures formed during degradation. It was observed that these metallic melts are able to attack the surrounding fuel rods, which could have consequences on fuel degradation and fission product release kinetics during severe accidents.

  8. 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.

  9. Attachment for sucker rod depth adjustment

    SciTech Connect

    Collins, N.D.

    1992-04-07

    This patent describes a surface unit of an oil well pumping system, having a walking beam, a suspended carrier bar and an interconnected sucker rod assembly for stroking a reciprocating down-hole pump. It comprises a cross bar having a centrally located passage therein for the sucker rod assembly and adapted to be transversely supported by the carrier bar; a depth adjusting bar, having a centrally located passage therein for the sucker rod assembly, positioned at a selected fixed dimension above and parallel to the cross bar and adapted to operatively support the sucker rod assembly; clamping means for fixing the sucker rod relative to the depth adjusting bar; and hydraulically extendable means supportively connecting the depth adjusting bar to the cross bar on at least each side of the carrier bar for adjusting the selected fixed dimension and maintaining the adjustment during operation.

  10. 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.

  11. Thermionic switched self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

    1989-01-01

    A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

  12. 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

  13. Quantum rods as nanocarriers of gene therapy.

    PubMed

    Aalinkeel, Ravikumar; Nair, Bindukumar; Reynolds, Jessica L; Sykes, Donald E; Law, Wing-Cheung; Mahajan, Supriya D; Prasad, Paras N; Schwartz, Stanley A

    2012-05-01

    Both antisense oligonucleotides (ASODN) and small interfering RNA (siRNA) have enormous potential to selectively silence specific cancer-related genes and could therefore be developed to be important therapeutic anti-cancer drugs. The use of nanotechnology may allow for significant advancement of the therapeutic potential of ASODN and siRNA, due to improved pharmacokinetics, bio-distribution and tissue specific targeted therapy. In this mini-review, we have discussed the advantages of using a nanocarrier such as a multimodal quantum rod (QR) complexed with siRNA for gene delivery. Comparisons are made between ASODN and siRNA therapeutic efficacies in the context of cancer and the enormous application potential of nanotechnology in oncotherapy is discussed. We have shown that a QR-interleukin-8 (IL-8) siRNA nanoplex can effectively silence IL-8 gene expression in the PC-3 prostate cancer cells with no significant toxicity. Thus, nanocarriers such as QRs can help translate the potent effects of ASODN/siRNA into a clinically viable anti-cancer therapy. Drug delivery for cancer therapy, with the aid of nanotechnology is one of the major translational aspects of nanomedicine, and efficient delivery of chemotherapy drugs and gene therapy drugs or their co-delivery continue to be a major focus of nanomedicine research.

  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. [Microecology of nuclear reactor pool water].

    PubMed

    Mal'tsev, V N; Saadavi, A; Aĭiad, A; El'gaui, O; Shlip, M

    1996-01-01

    In the course of research it was found that the circulation of pool water through the nuclear reactor core produces a bactericidal effect of microflora due to the influence of radiation of various types. The amount of microbes returns to initial level after 2-4 months after circulation was stopped. Microflora of pool water comprises large amounts of coccus, Gram-positive rods, fungi and a lower content of Gram-negative rods if compared to water which had been used to fill reactor pool. No difference in radioresistance was noticed for unitype microbes isolated from initial water and from reactor pool water. Quality of microflora reflects a unique phenomenon called "selection" which results in vanishing of all the radiosensitive types of microbes and survival of the radioresistant types. Radioresistance grows with increasing of catalase and nuclease activity.

  16. Quick release latch for reactor scram

    DOEpatents

    Johnson, M.L.; Shawver, B.M.

    1975-09-16

    A simple, reliable, and fast-acting means for releasing a control element and allowing it to be inserted rapidly into the core region of a nuclear reactor for scram purposes is described. A latch mechanism grips a coupling head on a nuclear control element to connect the control element to the control drive assembly. The latch mechanism is closed by tensioning a cable or rod with an actuator. The control element is released by de-energizing the actuator, providing fail-safe, rapid release of the control element to effect reactor shutdown. A sensing rod provides indication that the control element is properly positioned in the latch. Two embodiments are illustrated, one involving a collet- type latch mechanism, the other a pliers-type latch mechanism with the actuator located inside the reactor vessel. (auth)

  17. Quick release latch for reactor scram

    DOEpatents

    Johnson, Melvin L.; Shawver, Bruce M.

    1976-01-01

    A simple, reliable, and fast-acting means for releasing a control element and allowing it to be inserted rapidly into the core region of a nuclear reactor for scram purposes. A latch mechanism grips a coupling head on a nuclear control element to connect the control element to the control drive assembly. The latch mechanism is closed by tensioning a cable or rod with an actuator. The control element is released by de-energizing the actuator, providing fail-safe, rapid release of the control element to effect reactor shutdown. A sensing rod provides indication that the control element is properly positioned in the latch. Two embodiments are illustrated, one involving a collet-type latch mechanism, the other a pliers-type latch mechanism with the actuator located inside the reactor vessel.

  18. Enhancement of fiberglass sucker-rod design is offered

    SciTech Connect

    Hallden, D.F.

    1985-09-30

    This paper discribes the effective use of fiberglass-reinforced plastic sucker rods (FRP). FRP sucker rods have proven to be an economical solution to many sucker rod beam pumping problems. Two important characteristics that contribute to the effectiveness of FRP sucker rods are effective modulus of elasticity and fatigue life. Computerized simulations show that FRP sucker rod installations can benefit from using rod designs with a lower modulus of elasticity.

  19. CASL Structural Mechanics Modeling of Grid-to-Rod Fretting (GTRF)

    NASA Astrophysics Data System (ADS)

    Lu, Wei; Thouless, M. D.; Hu, Zupan; Wang, Hai; Ghelichi, Ramin; Wu, Chen-Hung; Kamrin, Ken; Parks, David

    2016-09-01

    Fluid-induced grid-to-rod fretting (GTRF) wear is responsible for over 70% of fuel leaks in pressurized water reactors (PWRs) in the US. The Consortium for Advanced Simulation of Light Water Reactors (CASL) has identified GTRF as a challenge problem that is very important to nuclear plants. GTRF is a complex problem that involves multiple physical phenomena. This paper summarizes several GTRF-related problems being addressed by the University of Michigan and the Massachusetts Institute of Technology, CASL partners. These include analyses of cladding creep, wear, structural mechanics, and the effects of the rod-to-grid gap. Also outlined are additional aspects of material science and computational modeling that will be needed to realize the ultimate goal of high-fidelity predictive modeling and design tools to address GTRF.

  20. Axial power monitor rod issues and resolution for K-14.1

    SciTech Connect

    Easterling, T.C.; Fields, C.C.; Hightower, N.T. III; Wooten, L.A.; Andre, S.K.; Apperson, C.E.; Bailey, M.H.; Bell, D.L.; Clare, G.H.; Collins, S.L.; Croft, W.D.; Edwards, T.B.; Geiger, G.T.; Harris, S.P.; Lietzow, J.L.; McCulloch, R.W.; McFarlane, A.F.; Randolph, H.W.; Reed, R.L.; Reeve, C.P.; Revolinski, S.M.; Sessions, H.T.; Shine, E.P.; Smith, T.A.; Sossman, C.L.; Taylor, J.J.; Weber, J.H.

    1992-05-01

    A recent concern arose over the treatment of uncertainty associated with the K-Reactor axial power monitors (APMs). There are nine axial power monitor rods located at various positions in the K-Reactor core. By comparing the output of one sensor near the top of the rod to the output of another sensor near the bottom of the rod, the relative ratio of the neutron flux from the top to the bottom of the core can be determined. This ratio is called the roof-top-ratio (RTR) and is the output of a top sensor (Sensor 2) divided by the output of a bottom sensor (Sensor 6). The RTR is important to the safety analyses because when the RTR is maintained within certain ranges, the severity of reactivity transients is limited. There are uncertainties associated with the equipment`s ability to measure the true roof top ratio. It was determined recently that sufficient uncertainty was not accounted for either in reactor operation or in the safety analyses. The concern about uncertainty was addressed for three separate issues. One issue dear with the linear response of the sensors for power ranges planned for K-Reactor operation. The second issue dear with overall uncertainty in the RTR channel. The third issue dear with apparent large ranges in confidence bands for the RTR at low reactor powers as represented by original vendor data. Plots of sparse vendor data indicated unacceptably large uncertainties in RTR would have to be accounted for at the power ranges planned for K-Reactor operation. These concerns were brought to management`s attention through the existing procedures for notification, irrespective of their potential impact on the restart schedule. Analyses have been completed to resolve the APM issues described above, and work is progressing to take the needed steps to change operational procedures.

  1. Determination of integral turbulence model parameters as applied to calculation of rod-bundle flows in porous-body approximation

    NASA Astrophysics Data System (ADS)

    Vlasov, M. N.; Korsun, A. S.; Maslov, Yu. A.; Merinov, I. G.; Rachkov, V. I.; Kharitonov, V. S.

    2016-03-01

    In the present paper, results of numerical simulation of single-phase flows of heat carrier through square and triangular rod bundles are reported. The simulations were aimed at the determination of parameters involved in an integral model of turbulence being developed for modeling nuclear-reactor cores and heat exchangers in anisotropic porous-body approximation.

  2. Dielectric Heaters for Testing Spacecraft Nuclear Reactors

    NASA Technical Reports Server (NTRS)

    Sims, William Herbert; Bitteker, Leo; Godfroy, Thomas

    2006-01-01

    A document proposes the development of radio-frequency-(RF)-driven dielectric heaters for non-nuclear thermal testing of the cores of nuclear-fission reactors for spacecraft. Like the electrical-resistance heaters used heretofore for such testing, the dielectric heaters would be inserted in the reactors in place of nuclear fuel rods. A typical heater according to the proposal would consist of a rod of lossy dielectric material sized and shaped like a fuel rod and containing an electrically conductive rod along its center line. Exploiting the dielectric loss mechanism that is usually considered a nuisance in other applications, an RF signal, typically at a frequency .50 MHz and an amplitude between 2 and 5 kV, would be applied to the central conductor to heat the dielectric material. The main advantage of the proposal is that the wiring needed for the RF dielectric heating would be simpler and easier to fabricate than is the wiring needed for resistance heating. In some applications, it might be possible to eliminate all heater wiring and, instead, beam the RF heating power into the dielectric rods from external antennas.

  3. Method speeds tapered rod design for directional well

    SciTech Connect

    Hu Yongquan; Yuan Xiangzhong

    1995-10-16

    Determination of the minimum rod diameter, from statistical relationships, can decrease the time needed for designing a sucker-rod string for a directional well. A tapered rod string design for a directional well is more complex than for a vertical well. Based on the theory of a continuous beam column, the rod string design in a directional well is a trial and error method. The key to reduce the time to obtain a solution is to rapidly determine the minimum rod diameter. This can be done with a statistical relationship. The paper describes sucker rods, design method, basic analysis rod design, and minimum rod diameter.

  4. 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

  5. Induced Current Measurement of Rod Vibrations

    NASA Astrophysics Data System (ADS)

    Sawicki, Charles A.

    2003-01-01

    The longitudinal normal modes of vibration of rods are similar to the modes seen in pipes open at both ends. A maximum of particle displacement exists at both ends and an integral number (n) of half wavelengths fit into the rod length. The frequencies fn of the normal modes is given by Eq. (1), where L is the rod length and V is the wave velocity: fn = nV/2L. Many methods have been used to measure the velocity of these waves. The Kundt's tube method commonly used in student labs will not be discussed here. A simpler related method has been described by Nicklin.2 Kluk3 measured velocities in a wide range of materials using a frequency counter and microphone to study sounds produced by impacts. Several earlier methods4,5 used phonograph cartridges complete with needles to detect vibrations in excited rods. A recent interesting experiment6 used wave-induced changes in magnetization produced in an iron rod by striking one end. The travel time, measured as the impulsive wave reflects back and forth, gave the wave velocity for the iron rod. In the method described here, a small magnet is attached to the rod with epoxy, and vibrations are detected using the current induced in a few loops of wire. The experiment is simple and yields very accurate velocity values.

  6. 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.

  7. Changing optical axis due to reactor operation

    NASA Astrophysics Data System (ADS)

    Hussey, D. S.; Jacobson, D. L.; Baltic, E.

    2011-09-01

    During reactor operation, the neutron flux distribution is modified by the reactor control mechanisms and in the case of the reactor at the National Institute of Standards and Technology, this is determined by the angular position of the Cd shim arms and the vertical position of an Al regulating rod. The changing flux distribution results in a change in the optical axis of neutron beams, whose view is a fixed position within the reactor core. The changing optical axis results in two noticeable image artifacts: poor registration between images of a static object taken at different times and a change in the shape of the flat field intensity. These two effects were measured during the first four days of reactor operation. Both measurements show correlation with the reactor control mechanisms, with combined correlation coefficients during the first two days after reactor startup approaching 1. The change in the edge position is well below the image spatial resolution, and has more uncertainty associated with it. However, the change in the flat-field shape demonstrates a clear correlation with both shim arm angle and regulating rod position.

  8. 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.

  9. SecA drives transmembrane insertion of RodZ, an unusual single-span membrane protein.

    PubMed

    Rawat, Swati; Zhu, Lu; Lindner, Eric; Dalbey, Ross E; White, Stephen H

    2015-03-13

    The transmembrane (TM) helices of most type II single-span membrane proteins (S-SMPs) of Escherichia coli occur near the N-terminus, where the cell's targeting mechanisms can readily identify it as it emerges from the ribosome. However, the TM helices of a few S-SMPs, such as RodZ, occur a hundred or more residues downstream from the N-terminus, which raises fundamental questions about targeting and assembly. Because of RodZ's novelty and potential usefulness for understanding TM helix insertion in vivo, we examined its membrane targeting and assembly. We used RodZ constructs containing immunotags before the TM domain to assess membrane insertion using proteinase K digestion. We confirmed the N(in)-C(out) (type II) topology of RodZ and established the absence of a targeting signal other than the TM domain. RodZ was not inserted into the membrane under SecA depletion conditions or in the presence of sodium azide, which is known to inhibit SecA. Insertion failed when the TM proton gradient was abolished with Carbonyl cyanide m-chlorophenyl hydrazone. Insertion also failed when RodZ was expressed in SecE-depleted E. coli, indicating that the SecYEG translocon is required for RodZ assembly. Protease accessibility assays of RodZ in other E. coli depletion strains revealed that insertion is independent of SecB, YidC, and SecD/F. Insertion was found to be only weakly dependent on the signal recognition particle pathway: insertion was weakly dependent on the Ffh but independent of FtsY. We conclude that membrane insertion of RodZ requires only the SecYEG translocon, the SecA ATPase motor, and the TM proton motive force.

  10. The Mechanical Effect of Rod Contouring on Rod-Screw System Strength in Spine Fixation

    PubMed Central

    Karakasli, Ahmet; Karaarslan, Ahmet A.; Ozcanhan, Mehmet Hilal; Ertem, Fatih; Erduran, Mehmet

    2016-01-01

    Objective Rod-screw fixation systems are widely used for spinal instrumentation. Although many biomechanical studies on rod-screw systems have been carried out, but the effects of rod contouring on the construct strength is still not very well defined in the literature. This work examines the mechanical impact of straight, 20° kyphotic, and 20° lordotic rod contouring on rod-screw fixation systems, by forming a corpectomy model. Methods The corpectomy groups were prepared using ultra-high molecular weight polyethylene samples. Non-destructive loads were applied during flexion/extension and torsion testing. Spine-loading conditions were simulated by load subjections of 100 N with a velocity of 5 mm min-1, to ensure 8.4-Nm moment. For torsional loading, the corpectomy models were subjected to rotational displacement of 0.5° s-1 to an end point of 5.0°, in a torsion testing machine. Results Under both flexion and extension loading conditions the stiffness values for the lordotic rod-screw system were the highest. Under torsional loading conditions, the lordotic rod-screw system exhibited the highest torsional rigidity. Conclusion We concluded that the lordotic rod-screw system was the most rigid among the systems tested and the risk of rod and screw failure is much higher in the kyphotic rod-screw systems. Further biomechanical studies should be attempted to compare between different rod kyphotic angles to minimize the kyphotic rod failure rate and to offer a more stable and rigid rod-screw construct models for surgical application in the kyphotic vertebrae. PMID:27651858

  11. The Mechanical Effect of Rod Contouring on Rod-Screw System Strength in Spine Fixation

    PubMed Central

    Karakasli, Ahmet; Karaarslan, Ahmet A.; Ozcanhan, Mehmet Hilal; Ertem, Fatih; Erduran, Mehmet

    2016-01-01

    Objective Rod-screw fixation systems are widely used for spinal instrumentation. Although many biomechanical studies on rod-screw systems have been carried out, but the effects of rod contouring on the construct strength is still not very well defined in the literature. This work examines the mechanical impact of straight, 20° kyphotic, and 20° lordotic rod contouring on rod-screw fixation systems, by forming a corpectomy model. Methods The corpectomy groups were prepared using ultra-high molecular weight polyethylene samples. Non-destructive loads were applied during flexion/extension and torsion testing. Spine-loading conditions were simulated by load subjections of 100 N with a velocity of 5 mm min-1, to ensure 8.4-Nm moment. For torsional loading, the corpectomy models were subjected to rotational displacement of 0.5° s-1 to an end point of 5.0°, in a torsion testing machine. Results Under both flexion and extension loading conditions the stiffness values for the lordotic rod-screw system were the highest. Under torsional loading conditions, the lordotic rod-screw system exhibited the highest torsional rigidity. Conclusion We concluded that the lordotic rod-screw system was the most rigid among the systems tested and the risk of rod and screw failure is much higher in the kyphotic rod-screw systems. Further biomechanical studies should be attempted to compare between different rod kyphotic angles to minimize the kyphotic rod failure rate and to offer a more stable and rigid rod-screw construct models for surgical application in the kyphotic vertebrae.

  12. 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.

  13. Taylor impact of glass rods

    NASA Astrophysics Data System (ADS)

    Willmott, G. R.; Radford, D. D.

    2005-05-01

    The deformation and fracture behavior of soda-lime and borosilicate glass rods was examined during classic and symmetric Taylor impact experiments for impact pressures to 4 and 10GPa, respectively. High-speed photography and piezoresistive gauges were used to measure the failure front velocities in both glasses, and for impact pressures below ˜2GPa the failure front velocity increases rapidly with increasing pressure. As the pressure was increased above ˜3GPa, the failure front velocities asymptotically approached maximum values between the longitudinal and shear wave velocities of each material; at ˜4GPa, the average failure front velocities were 4.7±0.5 and 4.6±0.5mmμs-1 for the soda-lime and borosilicate specimens, respectively. The observed mechanism of failure in these experiments involved continuous pressure-dependent nucleation and growth of microcracks behind the incident wave. As the impact pressure was increased, there was a decrease in the time to failure. The density of cracks within the failed region was material dependent, with the more open-structured borosilicate glass showing a larger fracture density.

  14. Taylor impact of glass rods

    SciTech Connect

    Willmott, G.R.; Radford, D.D.

    2005-05-01

    The deformation and fracture behavior of soda-lime and borosilicate glass rods was examined during classic and symmetric Taylor impact experiments for impact pressures to 4 and 10 GPa, respectively. High-speed photography and piezoresistive gauges were used to measure the failure front velocities in both glasses, and for impact pressures below {approx}2 GPa the failure front velocity increases rapidly with increasing pressure. As the pressure was increased above {approx}3 GPa, the failure front velocities asymptotically approached maximum values between the longitudinal and shear wave velocities of each material; at {approx}4 GPa, the average failure front velocities were 4.7{+-}0.5 and 4.6{+-}0.5 mm {mu}s{sup -1} for the soda-lime and borosilicate specimens, respectively. The observed mechanism of failure in these experiments involved continuous pressure-dependent nucleation and growth of microcracks behind the incident wave. As the impact pressure was increased, there was a decrease in the time to failure. The density of cracks within the failed region was material dependent, with the more open-structured borosilicate glass showing a larger fracture density.

  15. Neutron strain scanning in straightened eutectoid steel rods

    NASA Astrophysics Data System (ADS)

    Martínez, M. L.; Borlado, C. R.; Mompeán, F. J.; Peng, R. L.; Daymond, M. R.; Ruiz, J.; García-Hernández, M.

    Neutron strain scanning has been performed on a neutectoid steel rod at a reactor-based source (REST diffractometer, at NFL) and at a pulsed source (ENGIN diffractometer, at ISIS). The rod is primarily obtained from a drawing process and has been subject to bending and straightening procedures, which induce residual stress. The material exhibits a pearlitic microstructure, with alternating ferrite (90 vol%) and cementite (10 vol%) layers. Strain profiles for the ferritic phase were measured on REST. Both phases were measured on ENGIN and analysed by single-peak (ferrite) and Rietveld refinement (ferrite and cementite) methods. The agreement between REST and ENGIN data is excellent for the three measured directions in the ferritic phase. Total stress profiles have been evaluated by combining phase stresses using the rule of mixtures. The experimental results compare well with analytical models for a two-phase material subject to bending and straightening operations under pure bending and unbending moments with perfect elastic behaviour up to the yield point and plastic Voce behaviour above.

  16. 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.

  17. Method of cleaning and inhibiting sucker rod corrosion

    SciTech Connect

    Ford, M. B.; Griffin, J. B.

    1985-01-22

    Method of cleaning tubular goods, especially sucker rods, and inhibiting the sucker rods against corrosion as the rod string is being withdrawn from a borehole. The method is carried out by the provision of an enclosure which is attached to the upper end of a cased borehole. The upper end of the sucker rod string is extended axially through the enclosure as the rod string is withdrawn from the casing. A medial length of the rod string is engaged by a resilient packer device which wipes the rod clean of well fluids and loose debris. The rod string is next cleaned within a second chamber by impacting the outer surface thereof with an abrasive substance. The rod surface is again cleaned of any residual material. The rod is then moved through another chamber where corrosion inhibitor is applied to the external surface of the rod. As each treated joint of rod is withdrawn from the enclosure, the rod joints are sequentially unscrewed and suitably stacked, where the rods are protected from the elements, as well as being protected when the rods are subsequently made up into a rod string as the rod is replaced into a borehole.

  18. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    NASA Astrophysics Data System (ADS)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  19. 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.

  20. Neutron flux depression in the UO2-Pu2 (15 to .30%) fuel rods from IVO-FR2-Vg7-irradiation experiment

    NASA Astrophysics Data System (ADS)

    Lopezjiminez, J.; Fernandezmarron, J. L.

    The thermal-neutron flux depression within a fuel rod has a great influence in the radial temperature profile of the rod, especially for high enrichment fuel. The UO2-PuO2 (15 to 30% PuO2) fuel pins for the KfK-JEN joint irradiation program IVO were studied in the FR2 reactor. Different methods: (diffusion, Bonalumi, successive generations) were compared and a parabolic approximation approach was developed.

  1. TREAT Reactor Control and Protection System

    SciTech Connect

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.

  2. The attenuation of rod signals by bleachings

    PubMed Central

    Alpern, M.; Rushton, W. A. H.; Torii, S.

    1970-01-01

    1. Contrast flash technique allows the rod threshold to be measured even when it lies far above the cone threshold. In this way the rod dark adaptation curve after rhodopsin bleaching can be measured over 6 log units. 2. By retinal densitometry the regeneration of rhodopsin can be measured in the same subject. It is found that the log threshold is raised 1·2 units for each 10% of rhodopsin in the bleached state. 3. We have tried to discover whether bleaching raises the threshold by desensitizing the rods, or (like backgrounds) by attenuating their signals. Neither suggestion satisfies all conditions. 4. All are satisfied by [Formula: see text], where N is the size of rod signal, constant for threshold; θ, θD are steady backgrounds of light and receptor noise; ϕ is the threshold flash with σ a constant of about 2·5 log td sec; B the fraction of pigment in the bleached state. PMID:5499030

  3. 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.

  4. Reactor production of Thorium-229

    DOE PAGESBeta

    Boll, Rose Ann; Murphy, Karen E.; Denton, David L.; Tamara J. Haverlock; Garland, Marc A.; Mirzadeh, Saed; Hogle, Susan; Owens, Allison

    2016-05-03

    Limited availability of 229Th for clinical applications of 213Bi necessitates investigation of alternative production routes. In reactor production, 229Th is produced from neutron transmutation of 226Ra, 228Ra, 227Ac and 228Th. Here, we evaluate irradiations of 226Ra, 228Ra, and 227Ac targets at the ORNL High Flux Isotope Reactor.

  5. Reactor production of Thoruim-229

    DOE PAGESBeta

    Boll, Rose Ann; Murphy, Karen E.; Denton, David L.; Tamara J. Haverlock; Garland, Marc A.; Mirzadeh, Saed; Hogle, Susan; Owens, Allison

    2016-05-03

    Limited availability of 229Th for clinical applications of 213Bi necessitates investigation of alternative production routes. In reactor production, 229Th is produced from neutron transmutation of 226Ra, 228Ra, 227Ac and 228Th. Here, we evaluate irradiations of 226Ra, 228Ra, and 227Ac targets at the ORNL High Flux Isotope Reactor.

  6. Protein Kinase C Activity and Light Sensitivity of Single Amphibian Rods

    PubMed Central

    Xiong, W.-H.; Nakatani, K.; Ye, B.; Yau, K.-W.

    1997-01-01

    Biochemical experiments by others have indicated that protein kinase C activity is present in the rod outer segment, with potential or demonstrated targets including rhodopsin, transducin, cGMP-phosphodiesterase (PDE), guanylate cyclase, and arrestin, all of which are components of the phototransduction cascade. In particular, PKC phosphorylations of rhodopsin and the inhibitory subunit of PDE (PDE γ) have been studied in some detail, and suggested to have roles in downregulating the sensitivity of rod photoreceptors to light during illumination. We have examined this question under physiological conditions by recording from a single, dissociated salamander rod with a suction pipette while exposing its outer segment to the PKC activators phorbol-12-myristate,13-acetate (PMA) or phorbol-12,13-dibutyrate (PDBu), or to the PKC-inhibitor GF109203X. No significant effect of any of these agents on rod sensitivity was detected, whether in the absence or presence of a background light, or after a low bleach. These results suggest that PKC probably does not produce any acute downregulation of rod sensitivity as a mechanism of light adaptation, at least for isolated amphibian rods. PMID:9379174

  7. Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

    SciTech Connect

    Field, Kevin G.; Howard, Richard H.; Yamamoto, Yukinori

    2015-06-26

    The objective of the FeCrAl embrittlement screening tests being conducted through the use of Oak Ridge National Laboratories (ORNL) High Flux Isotope Reactor is to provide data on the radiation-induced changes in the mechanical properties including radiation-induced hardening and embrittlement through systematic testing and analysis. Data developed on the mechanical properties will be supported by extensive microstructural evaluations to assist in the development of structure-property relationships and provide a sound, fundamental understanding of the performance of FeCrAl alloys in intense neutron radiation fields. Data and analysis developed as part of this effort will be used to assist in the determination of FeCrAl alloys as a viable material for commercial light water reactor (LWR) applications with a primary focus as an accident tolerant cladding.

  8. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

    A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

  9. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  10. Calculator program speeds rod pump design

    SciTech Connect

    Engineer, R.; Davis, C.L.

    1984-02-01

    Matching sucker rod pump characteristics to a specific application is greatly simplified with this program, intended for use with an HP-41CV hand-held computer. The user inputs application data and the program calculates all necessary design criteria, including Mill's acceleration factor, peak and minimum polish rod loads and horsepower required. Sample calculations are provided, together with a thorough discussion of special design considerations involved in huff-and-puff applications.

  11. 1984 tubing and sucker rod tables

    SciTech Connect

    Not Available

    1984-01-01

    The first section of this handy reference lists companies that produce API tubing and couplings, giving specifications for pipe in sizes from 3/4 to 41/2 in. Also listed and illustrated are special tubing joints, identified by manufacturer Additional tables provide details on API sucker rods, including manufacturers, mechanical and chemical properties, dimensions and make-up recommendations. Similar data are presented for non-API rods.

  12. 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)

  13. 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.

  14. 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.

  15. Oil well sucker rod shock absorber

    SciTech Connect

    Knox, F.B.

    1986-02-18

    An oil well sucker rod shock absorber is described which consists of: an outer cylindrical casing defined by a cylindrical wall and having a removable upper plug and lower plug disposed respectively at upper and lower extremities of the casing. The upper plug has an axial bore and the lower plug defines a closed lower end and has an upwardly facing top surface. The plunger rod is connected to the sucker rod and is slidably disposed in the bore of the upper plug. A piston within the cylindrical casing is coupled to the plunger rod and has a downwardly facing bottom surface. Biasing means have a maximum vertical length disposed vertically within the casing and extending between the downwardly facing surface of the piston and the upwardly facing surface of the lower plug means at all times. This allows vertical reciprocal translation of the plunger rod and the piston within the cylindrical casing downwardly against the biasing means. Apertures are disposed through the cylindrical casing along the entire length thereof opposite the length of the biasing means, allowing downhole fluid pressure to be applied to the piston within the cylindrical casing via the apertures to be added to the force of the biasing means, without causing a fluid lock within the cylinder. Slap and wear of the sucker rod resulting therefrom are reduced and damage prevented.

  16. Crippling Strength of Axially Loaded Rods

    NASA Technical Reports Server (NTRS)

    Natalis, FR

    1921-01-01

    A new empirical formula was developed that holds good for any length and any material of a rod, and agrees well with the results of extensive strength tests. To facilitate calculations, three tables are included, giving the crippling load for solid and hollow sectioned wooden rods of different thickness and length, as well as for steel tubes manufactured according to the standards of Army Air Services Inspection. Further, a graphical method of calculation of the breaking load is derived in which a single curve is employed for determination of the allowable fiber stress. Finally, the theory is discussed of the elastic curve for a rod subject to compression, according to which no deflection occurs, and the apparent contradiction of this conclusion by test results is attributed to the fact that the rods under test are not perfectly straight, or that the wall thickness and the material are not uniform. Under the assumption of an eccentric rod having a slight initial bend according to a sine curve, a simple formula for the deflection is derived, which shows a surprising agreement with test results. From this a further formula is derived for the determination of the allowable load on an eccentric rod. The resulting relations are made clearer by means of a graphical representation of the relation of the moments of the outer and inner forces to the deflection.

  17. 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.

  18. 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.

  19. 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

  20. 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.

  1. 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.

  2. FLUID MODERATED REACTOR

    DOEpatents

    Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

    1957-10-22

    A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

  3. 28. A typical main control panel in a 105 reactor ...

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

    28. A typical main control panel in a 105 reactor building, in this case 105-F in February 1945. A single operator sat at the controls to regulate the pile's rate of reaction and monitor it for safety. The galvanometer screens (the two horizontal bars just below the nine round gauges that showed the positions of the control rods) showed the pile's current power setting. With that information, the operator could set the control rod positions to increase, decrease, or maintain the power. D-8310 - B Reactor, Richland, Benton County, WA

  4. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  5. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  6. Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

    SciTech Connect

    Higinbotham, W.A.

    1994-11-07

    The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 {+-} 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF.

  7. Apparatus and systems for measuring elongation of objects, methods of measuring, and reactor

    DOEpatents

    Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Condie, Keith G.; Stoots, Carl M.

    2011-11-29

    Elongation measurement apparatuses and systems comprise at least two Linear Variable Differential Transformers (LVDTs) with a push rod coupled to each of the at least two LVDTs at one longitudinal end thereof. At least one push rod extends to a base and is coupled thereto at an opposing longitudinal end, and at least one other push rod extends to a location spaced apart from the base and is configured to receive a sample between an opposing longitudinal end of the at least one other push rod and the base. Nuclear reactors comprising such apparatuses and systems and methods of measuring elongation of a material are also disclosed.

  8. 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.

  9. 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.

  10. Hanging core support system for a nuclear reactor

    DOEpatents

    Burelbach, James P.; Kann, William J.; Pan, Yen-Cheng; Saiveau, James G.; Seidensticker, Ralph W.

    1987-01-01

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform. Motion or radiation sensing detectors can be provide at the lower ends of the tension rods for obtaining pertinent readings proximate the core.

  11. Development of advanced strain diagnostic techniques for reactor environments.

    SciTech Connect

    Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

    2013-02-01

    The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

  12. Evaluation of the neutron background in an HPGe target for WIMP direct detection when using a reactor neutrino detector as a neutron veto system

    SciTech Connect

    Ji, Xiangpan; Xu, Ye Lin, Junsong; Feng, Yulong; Li, Haolin

    2013-11-15

    A direct WIMP (weakly interacting massive particle) detector with a neutron veto system is designed to better reject neutrons. The experimental configuration is studied in this paper involves 984 Ge modules placed inside a reactor-neutrino detector. The neutrino detector is used as a neutron veto device. The neutron background for the experimental design is estimated using the Geant4 simulation. The results show that the neutron background can decrease to O(0.01) events per year per tonne of high-purity germanium and it can be ignored in comparison with electron recoils.

  13. 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.

  14. Testing of inductively coupled Eddy current position sensor of diverse safety rod in sodium

    SciTech Connect

    Vijayashree, R.; Veeraswamy, R.; Nashine, B. K.; Dash, S. K.; Sharma, P.; Rajan, K. K.; Vijayakumar, G.; Rao, C. B.; Sosamma, S.; Kalyanasundaram, P.

    2011-07-01

    Prototype Fast Breeder Reactor (PFBR) is 500 MWe sodium cooled reactor under construction at Kalpakkam (India)). To improve the reliability of shutdown, Diverse Safety Rods (DSRs) are used in-addition to normal Control and Safety rods. During reactor operating condition, the DSR is parked above the active core and held in its top position by an electromagnet. In the event of a scram signal from the safety logic, the electromagnet holding the DSR is de-energised. Hence the DSR is released into the active core and at the end of travel DSR gets deposited in its bottom position. Because of the mechanical constraints, hard wired connectivity is not permitted from the DSR subassembly to the instrumentation outside the reactor. Hence an inductively coupled Eddy Current Position Sensor (ECPS) has been conceptualized to detect that the DSR has reached its bottom most position and to measure the drop time. Results of feasibility study on laboratory model have been reported earlier. Testing of a 1:1 scale engineering model of ECPS is reported in this paper. Results obtained from the high temperature sodium testing of ECPS indicate a clearly measurable change in pick up voltage with sensitivity of 11 % at 675 Hz. The ECPS is in advanced stage of implementation in DSRDM of PFBR. (authors)

  15. Reactor technology. Progress report, July-September 1980

    SciTech Connect

    Breslow, M.

    1980-12-01

    Progress in the Space Power Advanced Reactor (SPAR) Program includes indications that revision of the BeO reflector configuration can reduce system weight. Observed boiling limit restrictions on the performance of the annular-wick core heat pipe have accelerated transition to the development of the target-design arterial heat pipe. Successful bends of core heat pipes have been made with sodium as the mandrel material. With the phasing out of the GCFR Program, work on the Low Power Safety Experiments Program is now concentrated on completion of the third 37-rod Full Length Subgroup test. In the Reactor Safety/Structural Analysis area, effort on the Category I Structures Program is toward developing an experimental test plan focusing on a specific structural design. Buckling experiments on thin-walled cylindrical shells with circular cutouts are reported. Results of a three-dimensional analysis of thermal stresses in the Fort St. Vrain core support block are presented. Materials investigations and operation of a molybdenum-core SiC heat pipe are reported. Entrainment limits for gravity-assisted heat pipes and heat pipe configurations for application to energy conservation are being investigated. The new solution critical assembly, SHEBA, was completed. Godiva IV was temporarily relocated at TA-15. Influence of scattered radiations in the test vault on InRad measurements was determined from detector scans of the vault produced by /sup 252/Cf neutron and /sup 137/Cs gamma sources.

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

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

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

  17. End fitting for oil well sucker rods

    SciTech Connect

    Fischer, C.P.

    1984-02-07

    An end fitting for a sucker rod for oil wells is described with the end fitting having a chamber portion extending inwardly from one end thereof and an externally threaded portion at its other end. The chamber portion is defined by a plurality of spaced-apart annular ridges which define frusto-conical shaped cavities therebetween. The end fitting also has a bore extending inwardly thereinto from its other end which communicates with the inner end of the chamber portion. A valve is mounted in the end fitting and has a valve stem positioned in the bore and a valve head positioned at the inner end of the chamber portion. The chamber portion is adapted to receive a glass reinforced resin bonded cylindrical rod which is maintained therein by a two-part epoxy resin which surrounds the rod and is received in the cavities to form epoxy wedges bonded to the rod. The outer end of the bore is provided with internal threads which threadably receive a screw therein which engages the end of the valve stem so that longitudinal force may be applied to the valve thereby transmitting longitudinal force to the end of the rod.

  18. 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

  19. Dynamic behavior of rod photoreceptor disks.

    PubMed Central

    Chen, Chunhe; Jiang, Yunhai; Koutalos, Yiannis

    2002-01-01

    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

  20. MEANS FOR CONTROLLING A NUCLEAR REACTOR

    DOEpatents

    Wilson, V.C.; Overbeck, W.P.; Slotin, L.; Froman, D.K.

    1957-12-17

    This patent relates to nuclear reactors of the type using a solid neutron absorbing material as a means for controlling the reproduction ratio of the system and thereby the power output. Elongated rods of neutron absorbing material, such as boron steel for example, are adapted to be inserted and removed from the core of tae reactor by electronic motors and suitable drive means. The motors and drive means are controlled by means responsive to the neutron density, such as ionization chambers. The control system is designed to be responsive also to the rate of change in neutron density to automatically maintain the total power output at a substantially constant predetermined value. A safety rod means responsive to neutron density is also provided for keeping the power output below a predetermined maximum value at all times.

  1. NEUTRONIC REACTOR

    DOEpatents

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  2. Benchmark of CFD Simulations Using Temperatures Measured Within an Enclosed Array of Heater Rods Oriented Vertically and Horizontally

    NASA Astrophysics Data System (ADS)

    Chalasani, Narayana Rao

    Experiments and computational fluid dynamics/radiation heat transfer simulations of an 8x8 array of heated rods within an aluminum enclosure are performed. This configuration represents a region inside the channel of a spent boiling water reactor (BWR) fuel assembly between two consecutive spacer plates. The heater rods can be oriented horizontally or vertically to represent transport or storage conditions, respectively. The measured and simulated rod-to-wall temperature differences are compared for various heater rod power levels (100, 200, 300, 400 and 500W), gases (Helium and Nitrogen), enclosure wall temperatures, pressures (1, 2 and 3 atm) and orientations (Horizontal and Vertical) to assess the accuracy of the computational fluid dynamics (CFD) code. For analysis of spent nuclear fuel casks, it is crucial to predict the temperature of the hottest rods in an assembly to ensure that none of the fuel cladding exceeds its temperature limit. The measured temperatures are compared to those determined using CFD code to assess the adequacy of the computer code. Simulations show that temperature gradients are much steeper near the enclosure walls than they are near the center of the heater rod array. The measured maximum heater rod temperatures are above the center of heater rod array for nitrogen experiments in both horizontal and vertical orientations, whereas for helium the maximum temperatures are at the center of heater rod array irrespective of the orientation due to the high thermal conductivity of the helium gas. The measured temperatures of rods at symmetric locations are not identical, and the difference is larger for rods close to the enclosure wall than for those far from it. Small but uncontrolled deviations of the rod positions away from the design locations may cause these differences. For 2-inch insulated nitrogen experiment in vertical orientation with 1 atm pressure and a total heater rod power of 500 W, the maximum measured heater rod and enclosure

  3. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  4. Interpretation of calculated forces on sucker rods

    SciTech Connect

    Lea, J.F.; Pattillo, P.D. ); Studenmund, W.R. )

    1995-02-01

    The analysis of working loads in a sucker rod string during a pumping cycle has received substantial coverage in the petroleum literature. These load predictions have tended to focus on mechanical design considerations such as excess load and fatigue prediction. In contrast, the current study addresses the issues of buckling associated with working axial/pressure loads in an attempt to clarify the means of both predicting buckling and minimizing its effects. The study begins with a review of the static loads acting near the pump, and proceeds to a discussion of how these loads relate to the tendency of a rod string to buckle on the downstroke. Critical to this discussion is the concept of effective tension. Definition of the effective tension leads to the application of this concept to sinker bar design as a means of mitigating the buckling tendency of a rod string. Key points are reinforced by illustrative examples.

  5. System analysis for sucker-rod pumping

    SciTech Connect

    Schmidt, Z.; Doty, D.R.

    1989-05-01

    Pumping free gas in an oil well can significantly decrease the efficiency of a sucker-rod-pumping installation. Pump placement depth and use of a downhole gas/liquid separator (gas anchor) were found to be significant variables in improving the overall efficiency. A procedure is presented that shows when and to what degree the use of a gas anchor improves the efficiency of a sucker-rod pumping system. It was found that at lower pump intake pressures, the gas anchor usually improves efficiency, but at higher pump intake pressures, use of a gas anchor produces no positive effect. Also, elevating the pump to the highest position that still allows proper pump loading was found to reduce the operating costs of a sucker-rod-pumping installation significantly. Finally, a procedure is presented to calculate directly the pump volumetric efficiency and required volumetric pump displacement rate.

  6. System analysis for sucker rod pumping

    SciTech Connect

    Schmidt, Z.; Doty, D.R.

    1986-01-01

    Pumping free gas in an oil well can significantly decrease the efficiency of a sucker rod pumping installation. Pump placement depth and the use of a down hole gas-liquid separator (gas anchor) found to be significant variables in improving the overall efficiency. A procedure is presented which shows when and by how much the use of a gas anchor improves the efficiency of a sucker rod pumping system. It was found that at lower pump intake pressures the gas anchor usually improves efficiency, while at higher pump intake pressures the use of a gas anchor will produce no positive effect. Also, it was found at elevating the pump to the highest position which still allows for proper pump loading can significantly reduce the operating costs for a sucker rod pumping installation. Finally, a procedure is presented for directly calculating pump volumetric efficiency as well as the required volumetric pump displacement rate.

  7. Two-Phase Flow Patterns in a Four by Four Rod Bundle

    SciTech Connect

    Yoshitaka Mizutani; Shigeo Hosokawa; Akio Tomiyama

    2006-07-01

    Air-water two-phase flow patterns in a four by four square lattice rod bundle consisting of an acrylic channel box of 68 mm in width and transparent rods of 12 mm in diameter were observed by utilizing a high speed video camera, FEP (fluorinated ethylene propylene) tubes for rods, and a fiber-scope inserted in a rod. The FEP possesses the same refractive index as water, and thereby, whole flow patterns in the bundle and local flow patterns in subchannels were successfully visualized with little optical distortion. The ranges of liquid and gas volume fluxes, and , in the present experiments were 0.1 < < 2.0 m/s and 0.04 < < 8.85 m/s, which covered typical two-phase flow patterns appearing in a fuel bundle of a boiling water nuclear reactor. As a result, the following conclusions were obtained: (1) the region of slug flow in the - flow pattern diagram is so narrow that it can be regarded as a boundary between bubbly and churn flows, (2) the boundary between bubbly and churn flows is close to the boundary between bubbly and slug flows of the Mishima and Ishii's flow pattern transition model, and (3) the boundary between churn and annular flows is well predicted by the Mishima and Ishii's model. (authors)

  8. Design of a target and moderator at the Los Alamos Spallation Radiation Effects Facility (LASREF) as a neutron source for fusion reactor materials development

    SciTech Connect

    Ferguson, P.D.; Mueller, G.E.; Sommer, W.F.; Farnum, E.H.

    1993-10-01

    The LASREF facility is located in the beam stop area at LAMPF. The neutron spectrum is fission-like with the addition of a 3% to 5% component with E > 20 MeV. The present study evaluates the limits on geometry and material selection that will maximize the neutron flux. MCNP and LAHET were used to predict the neutron flux and energy spectrum for a variety of geometries. The problem considers 760 MeV protons incident on tungsten. The resulting neutrons are multiplied in uranium through (n,xn) reactions. Calculations show that a neutron flux greater than 10{sup 19} n/m{sup 2}/s is achievable. The helium to dpa ratio and the transmutation product generation are calculated. These results are compared to expectations for the proposed DEMO fusion reactor and to FFTF.

  9. Dark Light, Rod Saturation, and the Absolute and Incremental Sensitivity of Mouse Cone Vision

    PubMed Central

    Naarendorp, Frank; Esdaille, Tricia M.; Banden, Serenity M.; Andrews-Labenski, John; Gross, Owen P.; Pugh, Edward N.

    2012-01-01

    Visual thresholds of mice for the detection of small, brief targets were measured with a novel behavioral methodology in the dark and in the presence of adapting lights spanning ∼8 log10 units of intensity. To help dissect the contributions of rod and cone pathways, both wild-type mice and mice lacking rod (Gnat1−/−) or cone (Gnat2cpfl3) function were studied. Overall, the visual sensitivity of mice was found to be remarkably similar to that of the human peripheral retina. Rod absolute threshold corresponded to 12-15 isomerized pigment molecules (R*) in image fields of 800 to 3000 rods. Rod “dark light” (intrinsic retinal noise in darkness) corresponded to that estimated previously from single-cell recordings, 0.012R*s−1rod−1, indicating that spontaneous thermalisomerizations are responsible. Psychophysical rod saturation was measured for the first time in a nonhman species and found to be very similar to that of the human rod monochromat. Cone threshold corresponded to ∼5 R* cone−1 in an image field of 280 cones. Cone dark light was equivalent to ∼5000 R*s−1 cone−1, consistent with primate single-cell data but 100-fold higher than predicted by recent measurements of the rate of thermal isomerization of mouse cone opsins, indicating that nonopsin sources of noise determine cone threshold. The new, fully automated behavioral method is based on the ability of mice to learn to interrupt spontaneous wheel running on the presentation of a visual cue and provides an efficient and highly reliable means of examining visual function in naturally behaving normal and mutant mice. PMID:20844144

  10. Multigroup Reactor Lattice Cell Calculation

    1990-03-01

    The Winfrith Improved Multigroup Scheme (WIMS), is a general code for reactor lattice cell calculations on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters, and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered themore » choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are available in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a succesor version of WIMS-D/4.« less

  11. POWER GENERATING NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Vernon, H.C.

    1958-03-01

    This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

  12. A Special Topic From Nuclear Reactor Dynamics for the Undergraduate Physics Curriculum

    ERIC Educational Resources Information Center

    Sevenich, R. A.

    1977-01-01

    Presents an intuitive derivation of the point reactor equations followed by formulation of equations for inverse and direct kinetics which are readily programmed on a digital computer. Suggests several computer simulations involving the effect of control rod motion on reactor power. (MLH)

  13. 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.

  14. Pattern selection dynamics in rod eutectics

    NASA Astrophysics Data System (ADS)

    Serefoglu, Melis

    The cooperative or diffusively coupled growth of multiple phases during solidification is one of the most widely observed and generally important classes of phase transformations in materials. Technologically, low melting temperature and small freezing range contribute to excellent casting fluidity and fine composite structures give rise to favorable properties. Both of these features contribute to the wide application of eutectic alloys in the casting, welding, and soldering of engineered components. Despite the broad-based technological importance, many fundamental questions regarding eutectic solidification remain unanswered, severely limiting our ability to employ computational methods in the prediction of microstructure for the effective design of new materials and processes. At the core of the most persistent questions, lie problems involving multicomponent thermodynamics, solid-liquid and solid-solid interfacial phenomena, morphological stability, chemical and thermal diffusion, and nucleation phenomena. In the current study, pattern selection dynamics in rod eutectics are investigated using systematic directional solidification experiments and phase field simulations. Directional solidification of a succinonitrile-camphor (SCN-DC) transparent alloy in thin slab geometries of various thicknesses reveals two main points. First, a velocity is indentified at which a transition in array basis vectors is observed in specimens with many rows of rods (i.e. bulk). This transition amounts to a 90 degree rotation of the rod array, shifting from alignment of 1st nearest neighbors to alignment of 2nd nearest neighbors along the slide wall. Second, significant array distortion is observed with decreasing slide thickness, delta, which ultimately leads to a single-row (quasi-3D) morphology where delta/lambda is on the order of unity. In our analysis of these observations, we use a geometrical model to describe the rod arrangement as a function of slide thickness, providing

  15. Chemical Dosimeter Tube With Coaxial Sensing Rod

    NASA Technical Reports Server (NTRS)

    Lueck, Dale E.

    1993-01-01

    Improved length-of-stain (LOS) chemical dosimeter indicates total dose of chemical vapor in air. Made with rods and tubes of various diameters to obtain various sensitivities and dynamic ranges. Sensitivity larger and dose range smaller when more room for diffusion in gap between tube and rod. Offers greater resistance to changing of color of exposed dye back to color of unexposed condition, greater sensitivity, and higher degree of repeatability. Developed to measure doses of gaseous HCI, dosimeter modified by use of other dyes to indicate doses of other chemical vapors.

  16. Energy distributions in rods and beams

    NASA Technical Reports Server (NTRS)

    Wohlever, J. C.; Bernhard, R. J.

    1989-01-01

    A hypothesis proposed by Nefske and Sung (1987) that the mechanical energy flow in acoustic/structural systems can be modeled using a thermal energy flow analogy was tested for both longitudinal vibration in rods and transverse flexural vibrations in beams. It was found that the rod behaves according to the energy flow analogy. However, the beam solutions behaved significantly differently than predicted by the thermal analogy, unless spatially averaged energy and power flow were considered. Otherwise, the beam analysis is restricted to frequencies where the near-field terms in the displacement solution are negligible over most of the beam.

  17. Intraoperative pulmonary embolism of Harrington rod during spinal surgery: the potential dangers of rod cutting.

    PubMed

    Aylott, Caspar E W; Hassan, Kamran; McNally, Donal; Webb, John K

    2006-12-01

    This is a case report and laboratory-based biomechanics study. The objective is to report the first case of Titanium rod embolisation during scoliosis surgery into the Pulmonary artery. To investigate the potential of an unconstrained cut Titanium rod fragment to cause wounding with reference to recognised weapons. Embolisation of a foreign body to the heart is rare. Bullet embolisation to the heart and lungs is infrequently reported in the last 80 years. Iatrogenic cases of foreign body embolisation are very rare. Fifty 1-2 cm segments of Titanium rod were cut in an unconstrained manner and a novel method was used to calculate velocity. A high-speed camera (6,000 frames/s) was used to further measure velocity and study projectile motion. The wounding potential was investigated using lambs liver, high-speed photography and local dissection. Rod velocities were measured in excess of 23 m s(-1). Rods were seen to tumble end-over-end with a maximum speed of 560 revolutions/s. The maximum kinetic energy was 0.61 J which is approximately 2% that of a crossbow. This is sufficient to cause significant liver damage. The degree of surface damage and internal disruption was influenced by the orientation of the rod fragment at impact. An unconstrained cut segment of a Titanium rod has a significant potential to wound. Precautions should be taken to avoid this potentially disastrous but preventable complication.

  18. 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.

  19. Power Distribution Analysis for the ORNL High Flux Isotope Reactor Critical Experiment 3

    SciTech Connect

    Chandler, David; Primm, Trent; Maldonado, G Ivan

    2010-01-01

    The mission of the Reduced Enrichment for Research and Test Reactors Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low-enriched uranium (LEU) fuel and targets. Oak Ridge National Laboratory (ORNL) is currently reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction offlux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. Power distributions were calculated in the fuel elements of the HFIR, a research reactor at ORNL, via MCNP and were compared to experimentally obtained data. This study was performed to validate Monte Carlo methods for power density calculations and to observe biases. A current three-dimensional MCNP model was modified to replicate the 1965 HFIR Critical Experiment 3 (HFIRCE-3). In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. 'Foils' (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation, and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil's activity to the activity of a normalizing foil. This analysis consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the reference foil

  20. A neuronal circuit for colour vision based on rod-cone opponency.

    PubMed

    Joesch, Maximilian; Meister, Markus

    2016-04-14

    In bright light, cone-photoreceptors are active and colour vision derives from a comparison of signals in cones with different visual pigments. This comparison begins in the retina, where certain retinal ganglion cells have 'colour-opponent' visual responses-excited by light of one colour and suppressed by another colour. In dim light, rod-photoreceptors are active, but colour vision is impossible because they all use the same visual pigment. Instead, the rod signals are thought to splice into retinal circuits at various points, in synergy with the cone signals. Here we report a new circuit for colour vision that challenges these expectations. A genetically identified type of mouse retinal ganglion cell called JAMB (J-RGC), was found to have colour-opponent responses, OFF to ultraviolet (UV) light and ON to green light. Although the mouse retina contains a green-sensitive cone, the ON response instead originates in rods. Rods and cones both contribute to the response over several decades of light intensity. Remarkably, the rod signal in this circuit is antagonistic to that from cones. For rodents, this UV-green channel may play a role in social communication, as suggested by spectral measurements from the environment. In the human retina, all of the components for this circuit exist as well, and its function can explain certain experiences of colour in dim lights, such as a 'blue shift' in twilight. The discovery of this genetically defined pathway will enable new targeted studies of colour processing in the brain.

  1. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  2. Reactor building

    SciTech Connect

    Hista, J. C.

    1984-09-18

    Reactor building comprising a vessel shaft anchored in a slab which is peripherally locked. This reactor building comprises a confinement enclosure within which are positioned internal structures constituted by an internal structure floor, a vessel shaft, a slab being positioned between the general floor and the internal structure floor, the vesse

  3. Design of the Testing Set-up for a Nuclear Fuel Rod by Neutron Radiography at CARR

    NASA Astrophysics Data System (ADS)

    Wei, Guohai; Han, Songbai; Wang, Hongli; Hao, Lijie; Wu, Meimei; He, Linfeng; Wang, Yu; Liu, Yuntao; Sun, Kai; Chen, Dongfeng

    In this paper, an experimental set-up dedicated to non-destructively test a 15cm-long Pressurized Water Reactor (PWR) nuclear fuel rod by neutron radiography (NR) is described. It consists of three parts: transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo Simulation by the MCNP code. The material for the shell of the transport container was chosen to be lead with the thickness of 13 cm. Also, the mechanical devices were designed to control fuel rod movement inside the container. The imaging block was designed as the exposure platform, with three openings for the neutron beam, neutron converter foil, and specimen. Development and application of this experimental set-up will help gain much experience for investigating the actual irradiated fuel rod by neutron radiography at CARR in the future.

  4. 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.

  5. 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.

  6. 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.

  7. CADMIUM-RARE EARTH BORATE GLASS AS REACTOR CONTROL MATERIAL

    DOEpatents

    Ploetz, G.L.; Ray, W.E.

    1958-11-01

    A reactor control rod fabricated from a cadmiumrare earth-borate glass is presented. The rare earth component of this glass is selected from among those rare earths having large neutron capture cross sections, such as samarium, gadolinium or europium. Partlcles of this glass are then dispersed in a metal matrix by standard powder metallurgy techniques.

  8. 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.

  9. LPT. EBOR reactor vessel in TAN 646. Pressure vessel head ...

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

    LPT. EBOR reactor vessel in TAN 646. Pressure vessel head being installed in vault. Refueling port extension (right) and control rod nozzles (center). Camera facing northwest. Photographer: Comiskey. Date: January 20, 1965. INEEL negative no. 65-241 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

  10. NEUTRONIC REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

    DOEpatents

    Finniston, H.M.; Plail, O.S.

    1961-01-24

    BS>A uranium body for use in a nuclear fission reactor is described. It has a homogeneous rod of uranium metal enclosed in an envelope of aluminum, wherein a thin metallic layer of higher melting point than aluminum and of relatively low competitive neutron absorption between the uranium and the aluminum is bonded to the uranium and to the aluminum of the sheath.

  11. Analysis of sucker rod and sinkerbar failures

    SciTech Connect

    Waggoner, J.R.; Buchheit, R.G.

    1993-03-01

    This report presents results of a study of performance and failures of the sucker rod/sinkerbar string used in beam-pumping operations through metallography, finite element analysis, and failure data collection. Metallography showed that the microstructure of the steel bar stock needs to be considered to improve the fatigue resistance of the sucker rod strings. The current specification based on tensile strength, or yield strength, may not be appropriate since failure occurs because of fatigue and not yielding, and tensile strength is not always a good measure of fatigue resistance. Finite element analysis of the threaded connection quantitatively assesses the coupling designs under various loading conditions. Subcritical fractures in metallography are also suggested by calculated stress distribution in threaded coupling. Failure data illustrates both magnitude and frequency of failures, as well as categorizing the suspected cause of failure. Application of the results in each of these project areas is expected to yield improved choice of metal bar stock, thread design, and make-up practices which can significantly reduce the frequency of sucker rod failures. Sucker rod failures today are not inherent in the process, but can be minimized through the application of new technology and observation of common-sense practices.

  12. Program optimizes sucker-rod pumping mode

    SciTech Connect

    Takacs, G. )

    1990-10-01

    Direct energy costs for sucker-rod pumping can be optimized by selecting the right pump size, stroke length, and pumping speed for the required liquid production rate. Calculation procedures for a computer program are developed for optimizing the design of conventional pumping units.

  13. Stop sucker rod failures to save money

    SciTech Connect

    Moore, K.H.

    1981-07-01

    This study presents examples of frequent and common sucker rod failures, explains how failures occur, presents methods to recognize these failures, and discusses changes in conditions that cause failure. From early identification, corrective measures can be taken to prevent their recurrence, reducing downtime and lost production.

  14. Method of making class D sucker rods

    SciTech Connect

    Woodings, R. T.

    1984-12-04

    It has been found that API Class D sucker rods can be made inexpensively from low-alloy, low-cost steel by following a suitable induction-normalizing process and using a suitable steel to which there has been added 0.07 to 0.15 percent of vanadium.

  15. Wear simulation of sucker rod couplings

    SciTech Connect

    Schumacher, W.J. )

    1991-09-01

    This paper reports that sucker rod strings are devices used to actuate pumps located at the bottom of oil wells. The individual rods are connected together by threaded couplings. Since the couplings have a larger diameter than the rods, they sometimes contact the inside diameter of the tubing during the up and down pumping cycle. Usually, this is not problem unless buckling occurs in the downstroke; however, this can lead to accelerated wear of the coupling and tubing. In nonvertical wells (offset, deviated, or slanted), the contact is more severe and rapid wear takes place. Couplings are more easily replaced during shutdowns; it is very important to minimize wear to tubing since it is virtually impossible to replace. TRIBONIC 20, an iron-based alloy containing approximately 13% Mn, 5% Si, 5.5% Cr, and 5% Ni, was laboratory evaluated to determine whether or not it could solve the sucker rod coupling-production tubing wear problem. The alloy demonstrated outstanding wear resistance both to itself and in protecting type 1019 steel.

  16. 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.

  17. 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.

  18. 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

  19. 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

  20. Rod Soltis: Making Connections. Appalachian Scene.

    ERIC Educational Resources Information Center

    Baldwin, Fred D.

    1998-01-01

    Describes the work of Rod Soltis in developing interlinked telecommunications networks in all 14 of New York's Appalachian counties. The networks connect to each other, state and federal agencies and networks, schools, social service agencies, hospitals, and museums, and include private partnerships with telephone and cable TV companies. Soltis'…